Operational Radiation Safety Program: Recommendations of the National Council on Radiation Protection and Measurements (N C R P Report)

NCRP REPORT No. 127

OPERATIONAL RADIATION SAFETY PROGRAM Recommendations of the NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS

Issued June 12,1998

National Council on Radiation Protection and Measurement 7910 Woodmont Avenue / Bethesda, Maryland 20814-3095

LEGAL NOTICE This Report was prepared by the National Council on Radiation Protection and Measurements (NCRP). The Council strives to provide accurate, complete and useful information in its documents. However, neither the 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 thc 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. a s amended 42 U.S.C. Section 2000e et seq. (Title VII) or any other statutory or common law theorygoverning liability.

Library of Congress Cataloging-in-PublicationData National Council on Radiation Protection and Measurements. Operational radiation safety program : recommendations of the National Council on Radiation Protection and Measurements. p. cm. -- (NCRP report ; no. 127) "Issued June 1998." Includes bibliographical references and index. ISBN 0-929600-59-2 1. Radiation-Safety measures. I. National Council on Radiation Protection and Measurements. 11. Series TK9152.063 1998 363.17'996--dc21 98-4407 CIP

Copyright O National Council on Radiation Protection and Measurements 1998 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 copyrightowner, except for brief quotation in critical articles or reviews.

Preface NCRP Report No. 59, Operational Radiation Safety Program, was published in 1978. That report provided the philosophy, basic principles and requirements for a radiation safety program. In the intervening years, there have been many new developments including: new NCRP recommendations for limiting exposure to ionizing radiation (NCRP Report No. 91 in 1987 which was superseded by NCRP Report No. 116 in 1993); new techniques for the measurement and control of exposures and the disposal of radioactive waste; and new applications for ionizing radiation and radioactive materials. These developments served as the Council's rationale for preparing the current Report which supersedes NCRP Report No. 59. This Report reiterates the basic principles for establishing and maintaining an effective operational radiation safety program. Relevant aspects of such a program are discussed including: facility design criteria, organizationaVmanagementissues, training, internal and external radiation control strategies, radioactive waste disposal, environmental monitoring, radiation safety instrumentation, and emergency response planning. This Report does not attempt to summarize the regulatory or licensing requirements of the various federal, state or local authorities that may have jurisdiction over matters addressed in this publication. This Report was prepared by NCRP Scientific Committee 46. Serving on the Committee were:

Kenneth R. Kase, Chairman (1991-1 Stanford Linear Accelerator Center Menlo Park, California Members

John W.Baum (1993-) Brookhaven National Laboratory Upton, New York

Kenneth L. Miller (1995-) M.S. Hershey Medical Center Hershey, Pennsylvania

iv / PREFACE

Joyce P. Davis (1991-) Defense Nuclear Facilities Safety Board Washington, D.C.

David S. Myers (1991-1 Lawrence Livermore National Laboratory Livermore, California

Steven M. Garry (1996-) Florida Power Corporation Crystal River, Florida

J o h n W. Poston, Sr. (1991-1 Texas A&M University College Station, Texas

Duane C. Hall (1995-) 3M Health Physics Services St. Paul, Minnesota

Keith Schiager (1991-1997) Salt Lake City, Utah

William R. Hendee (1991-1995) Medical College of Wisconsin Milwaukee, Wisconsin

Ralph H. Thomas (1991-1996) Moraga, California

Kathryn A. Higley (1997-) Oregon State University Corvallis, Oregon

Paul G. Voillequk (1993-) M J P Risk Assessment, Inc. Idaho Falls, Idaho

Susan M. Langhorst (1995) University of MissouriColumbia Columbia, Missouri

Robert G. wissink* (1991-1995) 3M Health Physics Services St. Paul, M i ~ e S o t a

James E. ~ c ~ a u ~ h l i n * (1991-1995) Sante Fe, New Mexico

NCRP Secretariat Eric E. Kearsley (1997-), Staff Scientist Thomas M . Koval(1993-1997), Senior Staff Scientist J a m e s A. Spahn, Jr. (1991-1993), Senior Staffscientist Cindy L. O'Brien, Editorial Assistant

PREFACE / V

The Council wishes to express its appreciation to the Committee members for the time and effort devoted to the preparation of this Report. The Council also gratefdly acknowledges the support provided by the Health Physics Society in 1998 that permitted the completion of this Report.

Charles B. Meinhold President

Contents ... F'reface ............................................. 111 1 Introduction ..................................... 1 1.1 Purpose of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 Purpose of the Operational Radiation Safety Program ..................................... 2 2 Application of ALARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.1 Applicability of Cost-Benefit Analysis in the ALARA Process ............................... 7 2.2 Concepts of a Cost-Benefit Approach to ALARA . . . . . 8 2.2.1 Applicability of Collective Effective Dose . . . . . 8 2.23 Dose Magnitude and Distributions . . . . . . . . . . 8 2.2.3 Monetary Value of Avoided Dose ........... 9 2.3 Screening for ALARA Assessment . . . . . . . . . . . . . . . 11 3 Organization and Administration . . . . . . . . . . . . . . . . . 12 3.1. Management Commitment and Policy . . . . . . . . . . . . 12 3.2 Radiation Safety Organization . . . . . . . . . . . . . . . . . . 13 3.2.1 Radiation Safety Advisory Organization . . . . 13 3.2.2 Radiation Safety Officer . . . . . . . . . . . . . . . . . 14 3.3 Accreditation and Certification ................. 14 3.4 Radiation Safety Program Policies and Procedures .................................. 15 3.4.1 Radiation Safety Manual ................. 15 3.4.2 Radiation Safety Operating Procedures . . . . . 16 3.5 Responsibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.6 Quality Assurance ............................ 18 3.6.1 Management Goals ..................... 19 3.6.2 Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.6.3 Program Audits ........................ 20 3.6.4 Incident and Accident Investigations ....... 21 3.6.5 Deficiency Tracking . . . . . . . . . . . . . . . . . . . . . 22 3.7 Records Management .........................23 3.8 Occupational Medicine ........................ 23 3.9 Recommended Additional Reading ............... 24 4 Facility Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 4.1 Site Selection ................................ 26 4.2 Facility Layout ...............................28

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viii / CONTENTS Equipment and System Design .................. 29 Shielding ....................................30 Ventilation .................................. 32 Radioactive Material Waste Management ......... 35 Instrumentation and Access Control Systems ......36 Nuclear Criticality Safety ...................... 36 Recommended Additional Reading ............... 36 5 Orientation and Training ........................ 38 5.1 General Principles ............................ 38 5.2 Design of a General Training Program ............ 39 5.3 Specific Training Requirements ................. 41 6 External Radiation Exposure Control . . . . . . . . . . . . .42 6.1 Radiation Dose Controls . . . . . . . . . . . . . . . . . . . . . . . 43 6.1.1 Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 6.1.2 Administrative Dose Guidelines . . . . . . . . . . .43' 6.2 Radiation Dose Control Techniques . . . . . . . . . . . . . .43 6.2.1 Time. Distance and Shielding . . . . . . . . . . . . . 44 6.2.2 Access Control and Alarm Systems . . . . . . . . . 45 6.2.3 Radiation Safety Procedures and Radiation Work Permits . . . . . . . . . . . . . . . . . . . . . . . . . .48 6.2.4 Exposure Planning and Dose Reduction Activities .............................. 49 6.3 External Radiation Dosimetry .................. 49 6.3.1 Personal Monitoring ..................... 49 6.3.2 Dose Assessment ....................... 51 6.4 Monitoring and Surveillance Program ............51 6.4.1 Radiation Surveys ...................... 51 6.4.2 Area Monitoring ........................ 53 6.5 Protective Clothing ........................... 53 6.6 Records ..................................... 54 6.7 Recommended Additional Reading ............... 55 7 Internal Radiation Exposure Control .............56 7.1 Radiation Dose Controls ....................... 57 7.1.1 Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .57 7.1.2 Administrative Exposure Guidelines and Reference Levels . . . . . . . . . . . . . . . . . . . . . . . .57 7.2 Contamination Control Programs . . . . . . . . . . . . . . . .57 7.2.1 Access Control and Alarm Systems ......... 59 7.2.2 Radiation Safety Procedures and Radiation Work Permits . . . . . . . . . . . . . . . . . . . . . . . . . .60 7.2.3 Exposure Planning and Dose Reduction Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 7.3 Internal Radiation Dosimetry . . . . . . . . . . . . . . . . . . .62

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4.3 4.4 4.5 4.6 4.7 4.8 4.9

CONTENTS / i~

7.3.1 Personal Monitoring . . . . . . . . . . . . . . . . . . . . 62 7.3.2 Bioassay Measurements . . . . . . . . . . . . . . . . .63 7.3.3 Dose Assessment . . . . . . . . . . . . . . . . . . . . . . .6 5

Monitoring and Surveillance Program . . . . . . . . . . . . 66 7.4.1 Monitoring for Airborne Radioactivity . . . . . . 66 7.4.2 Contamination Surveys . . . . . . . . . . . . . . . . . . 69 7.5 Protective Equipment and Devices . . . . . . . . . . . . . . . 69 7.5.1 Containment Systems . . . . . . . . . . . . . . . . . . .69 7.5.2 Respiratory Protection . . . . . . . . . . . . . . . . . . . 70 7.5.3 Protective Clothing . . . . . . . . . . . . . . . . . . . . . 70 7.6 Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7. 1 8. Control of Low-Level Radioactive Waste . . . . . . . . . . 73 8.1 Minimizing the Production of Waste . . . . . . . . . . . . . 74 8.1.1 Practices for Minimizing Waste . . . . . . . . . . . 74 8.1.2 Practices for Reducing Mixed Waste ....... 75 8.2 Decontamination and Reuse of Tools and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 8.3 Collecting. Sorting and Classifying Waste . . . . . . . . . 76 8.4 Radioactive Waste Volume Reduction . . . . . . . . . . . . 77 8.5 Storage o f w a s t e . . . . . . . . . . . . . . . . . . . . . . . . . . . . .78 8.6 Disposal of Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . .78 8.7 Recycling of Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . .79 8.8 Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .79 8.9 Recommended Additional Reading . . . . . . . . . . . . . . . 80 9. Control of Exposure to the Public . . . . . . . . . . . . . . . . 81 9.1 Standards and Guidance . . . . . . . . . . . . . . . . . . . . . . . 81 9.2 Control of Off-Site Exposures . . . . . . . . . . . . . . . . . . . 82 9.2.1 Determining the Need for Monitoring . . . . . . 8 3 9.2.2 Monitoring Airborne Effluents . . . . . . . . . . . .84 9.2.3 Monitoring Liquid Effluents . . . . . . . . . . . . . . 86 9.2.4 Monitoring Solid Waste . . . . . . . . . . . . . . . . . . 86 9.3 Environmental Monitoring . . . . . . . . . . . . . . . . . . . . .87 9.3.1 Preoperational Monitoring . . . . . . . . . . . . . . . 88 9.3.2 Operational Monitoring . . . . . . . . . . . . . . . . . . 89 9.4 Measurement Methods . . . . . . . . . . . . . . . . . . . . . . . . 90 9.5 Dose Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 1 9.6 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . .92 9.7 Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 10 Radiation Safety Instrumentation . . . . . . . . . . . . . . . . 94 10.1 Instrument Specification . . . . . . . . . . . . . . . . . . . . . . .95 102 Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 10.3 Instrument Maintenance . . . . . . . . . . . . . . . . . . . . . . 98 10.4 Use of Instruments and Acceptable Uncertainty . . . 99 7.4

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CONTENTS

10.5 Selection of Instruments for Various Applications . 100 10.6 Records for an Instrument Program . . . . . . . . . . . . .106 10.7 Recommended Additional Reading . . . . . . . . . . . . . . 107 11 Planning for Radiation Emergencies . . . . . . . . . . . . .108 11.1 Development of the Emergency Plan . . . . . . . . . . . .108 11.2 Preparation of Implementing Procedures . . . . . . . . . 109 11.3 Classification of Emergencies . . . . . . . . . . . . . . . . . .110 11.4 Practical Considerations . . . . . . . . . . . . . . . . . . . . . . 111 11.5 Evaluation of the Plan . . . . . . . . . . . . . . . . . . . . . . . . 112 Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .114 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 TheNCRP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 NCRPPublications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .137 Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146

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1. Introduction 1.1 Purpose of this Report

In 1978, the National Council on Radiation Protection and Measurements (NCRP) published Report No. 59, Operational Radiation Safety Program (NCRP, 1978a) to provide, in a systematic way, the philosophy and the basic principles and requirements for a n operational radiation safety program. Since that time, a number of reports detailing specific aspects of operational radiation safety have been published by the Council. These include, NCRP Report No. 71, Operational Radiation Safety-Training (NCRP, 1983a); NCRP Report No. 88, Radiation Alarms and Access Control Systems (NCRP, 1986);NCRP Report No. 105,Radiation Protection for Medical and Allied Health Personnel (NCRP, 1989a); NCRP Report No. 107, Implementation of the Principle of As Low as Reasonably Achievable MLARA) for Medical and Dental Personnel (NCRP, 1990); NCRP Report No. 111, Developing Radiation Emergency Plans for Academic, Medical or Industrial Facilities (NCRP, 1991a); NCRP Report No. 112, Calibration of Survey Instruments Used i n Radiation Protection for the Assessment of Ionizing Radiation Fields and Radioactive Surface Contamination (NCRP, 1991b); NCRP Report No. 114, Maintaining Radiation Protection Records (NCRF', 1992); NCRP Report No. 118, Radiation Protection i n the Mineral Extraction Industry (NCRP, 1993a); NCRP Report NO. 120, Dose Control a t Nuclear Power Plants (NCRP, 1994); and NCRP Report No. 122, Use of Personal Monitors to Estimate Effective Dose Equivalent and Effective Dose to Workers for External Exposure to Low-LET Radiation (NCRP, 1995a). Reports in progress in the area of operational radiation safety include those on radiation safety design guidelines for particle accelerator facilities, assessment of occupational exposure from internally deposited radionuclides, radiation safety related to special medical procedures, and shielding design for radiotherapy facilities. Since the publication of NCRP Report No. 59 (NCRP, 1978a1, new recommendations have been made by the NCRP for limiting exposure to ionizing radiation (NCRP, 1993b). In addition, new applications for radiation and radioactive materials in research,

2 / 1. INTRODUCTION

medicine and industry have been developed. Techniques for the measurement and control of radiation exposure as well as the disposal of radioactive waste material have evolved. The principle that radiation exposures should be kept as low as reasonably achievable, economic and social factors being taken into account (the ALARA principle) now guides the development of operational radiation safety programs. The above factors provided the motivation to revise NCRP Report No. 59 (NCRP, 1978a). This Report is not intended to be a design manual, e.g., for radiation shielding or ventilation systems. Its objective is to describe the elements of a n operational radiation safety program that is based on the implementation of the ALARA principle below the radiation dose limits. Basic principles and practices of radiation safety are emphasized. Relevant elements of various NCRP reports pertaining to specific types of facilities or specific aspects of radiation safety are incorporated into the specifications provided here for operational radiation safety programs. This Report should provide guidance for the development of new radiation safety programs and serve as a useful tool for assessing mature radiation safety programs. For management personnel, this Report provides information about the basic requirements of a radiation safety program. It details specific aspects of operational radiation safety and references more detailed information in other NCRP reports, publications of the International Commission on Radiological Protection (ICRP), and other consensus bodies such as the American National Standards Institute (ANSI). This Report does not address regulatory or licensing requirements that may be imposed on a radiation protection program by state, local or federal authorities. 1.2 Purpose of the Operational Radiation Safety Program

Every institution and organization that uses nonexempt quantities of radioactive material or regulated devices that produce ionizing radiation should provide a program plan that specifies the policies and practices that are necessary to control radiation exposures to its employees and the public within the prescribed limits and to levels that are ALARA. The operational radiation safety program is the mechanism for the implementation of that plan. The size and definition of the program should be commensurate with the potential hazards.

1.2 PURPOSE OF THE OPERATIONAL RADIATION SAFETY PROGRAM

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The objective of a comprehensive radiation safety program is to protect people from the deleterious health effects that may result from exposure to ionizing radiation. Large radiation doses can cause such effects within a short time. Because such large doses, except for medical radiation therapy, are never intended, but are possible in the event of certain accidents, the radiation safety program should function to reduce the likelihood of accidents through careful facility and equipment design, safety procedures, and training (see Sections 3 , 4 and 5). Failures in facility design, failures in equipment, and human error can lead to unnecessary radiation exposure of individuals. Plans should be made and individuals should be trained for normal procedures as well as for emergencies (see Section 11).Even with the most careful planning and training, an accident (or near accident) can occur. Consequently, procedures should be established for evaluating failures, whether or not they result in accidents. The cause of any failure should be identified and actions should be taken to prevent recurrences. Normally, work with radiation sources does not result in radiation doses large enough to cause immediate or observable effects. However, the accumulation of radiation dose over a long period of time may result in an increased risk for delayed health effects. The NCRP recommends both annual and cumulative dose limits for individuals (see Table 1.1)that limit the risk to workers and the public (NCRP, 1993b). Program and facility design, and worker training are important to ensure that radiation exposures remain within these limits and are ALARA (see Sections 2, 3, 4 and 5). In addition, the program should include adequate control and evaluation of radiation exposures and radioactive wastes (see Sections 6, 7 , 8 and 9). Because radiation measurements are necessary for any radiation safety program, Section 10 provides information about the instrumentation that can be used for that purpose. Certain sections of this Report may not be applicable to a particular program. Consequently, there is some intentional redundancy included to remove interdependency between sections. This is especially true for Sections 6 and 7. In addition to the list of references supporting specific statements in the text of this Report (see page 119), five sections include lists of recommended additional reading. These lists are to be found at the end of Sections 3 , 4 , 6 , 8 and 10. A Glossary is also provided.

4 / 1. INTRODUCTION

TABLE1.l-Summary of NCRP recommendations specifying limits for radiation exposure [adapted from Table 19.1 of NCRP Report No. 116 (NCRP, 1993bll.a A. Occupational exposuresb 1. Effective dose limits a. Annual b. Cumulative

50 mSv 10 mSv

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age

2. Equivalent dose limits for tissues and organs (annual) a. Lens of eye b. Skin, hands and feet

B. Public exposures (annual) 1. Effective dose limit, continuous or frequent exposureb 2. Effective dose limit, infrequent exposureb 3. Equivalent dose limits for tissues and organsb a. Lens of eye b. Skin, hands and feet 4. Remedial action for natural sources a. Effective dose (excluding radon) b. Exposure to radon decay products C. Education and training exposures (annuaUb 1. Effective dose limit 2. Equivalent dose limits for tissues and organs a. Lens of eye b. Skin, hands and feet

D. Embrydfetus exposures (monthly)b 1.Equivalent dose limit E. Negligible individual dose per source or practice ( a n n ~ a l ) ~ a Excluding medical

exposures. Sum of internal and external exposures but excluding doses from natural sources.

2. Application of ALARA The basic radiation protection assumptions and objectives recommended by the Council are given in NCRP Report No. 116, Limitation of Exposure to Ionizing Radiation (NCRP, 1993b). Specifically: Based on the hypothesis that genetic effects and some cancers may result from damage to a single cell, the Council assumes that, for radiation-protection purposes, the risk of stochastic effects is proportional to dose without threshold, throughout the range of dose and dose rates of importance in routine radiation protection. Furthermore, the probability of response (risk) is assumed, for radiation-protection purposes, to accumulate linearly with dose. At higher doses, received acutely, such as in accidents, more complex (nonlinear) dose-risk relationships may apply. Given the above assumptions, radiation exposure a t any selected dose limit will, by definition, have an associated level of risk. For this reason, NCRP reiterates its previous recommendations concerning: (1)the need to justify any activity which involves radiation exposure on the basis that the expected benefits to society exceed the overall societal cost (justification), (2) the need to ensure that the total societal detriment from such justifiable activities or practices is maintained ALARA, economic and social factors being taken into account and (3) the need to apply individual dose limits to ensure that the procedures of justification and ALARA do not result in individuals or groups of individuals exceeding levels of acceptable risk (limitation). Justification is not normally a radiation protection consideration and the dose limits are now considered simply as upper bounds. As a result, the radiation protection program is driven primarily by

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2. APPLICATION OF ALARG

ALARA considerations. In most applications, ALARA is simply the continuation of good radiation protection programs and practices which have traditionally been effective in keeping the average of individual exposures of monitored workers well below the limits (NCRP, 1989b). Many of the decisions involved in control of radiation exposure result, primarily, from professional judgement of those responsible for health protection. Operationally, this is achieved by the application of good practices based on staff knowledge, training and, very frequently, common sense. In general, a graded approach is needed for making decisions based on the unusualness or complexity of the operation. For example, if the operation is routine and the potential for radiation exposure is small, only a small and inexpensive effort can be justified to avoid the exposure. Whereas, if the operation is new, and the potential for significant radiation dose is high, a much greater effort and expense can be justified. Most situations fall between these two extremes. Perhaps the most important approach to achieving ALARA is creating the proper "mind set" in managers, supervisors and workers so that they always ask if a particular level of exposure is necessary. In a well organized facility, almost all the technical decisions will have been made during planning and design. During operations there must be constant awareness and attention given to avoiding unnecessary exposures. Thorough work planning is a vital part of the ALARA process. Many times a small amount of shielding can be added to reduce the dose that workers might receive. Administrative controls on exposure can be used to identlfy work processes and procedures that may be modified to reduce exposures a t little cost. Three NCRP reports deal with the application of the ALARA principle in very different operational situations. NCRP Report No. 107, Implementation of the Principle of As LAW As Reasonably Achievable (ALARA) for Medical and Dental Personnel (NCRP, 1990) described its integration into radiation safety in medical and dental facilities. NCRP Report No. 120, Dose Control at Nuclear Power Plants (NCRP, 1994) discussed the use of the ALARA principle in dose control programs a t nuclear power plants. A third, NCRP Report No. 121, Principles and Application of Collective Dose in Radiation Protection (NCRP, 1995b), is closely related to the application of the ALARA principle. ICRP issued Publication 37 on ALARA, Cost-Benefit Analysis in the Optimization of Radiation Protection (ICRP, 1983). That publication stresses cost-benefit

2.1 APPLICABILITY OF COST-BENEFIT ANALYSIS / 7 approaches, while ICRP Publication 55, Optimization and Decision-Making in Radiological Protection (ICRP, 19891, suggests other approaches.

2.1 Applicability of Cost-Benefit Analysis in the ALARA Process

Instituting procedures for applying the ALARA principle will require the judgment of radiation safety professionals. When the potential for exposure of people to significant radiation doses exists, quantitative cost-benefit analyses may be justified to arrive a t the optimum approach for dose control. This Section presents the NCRP guidance for using some quantitative approaches that are important in applying the ALARA principle in the context of operational radiation safety. Protective measures that go beyond the basic design requirements should be considered and evaluated to determine the incremental cost related to the value of the collective effective dose avoided. Stated another way, the incremental cost of any elective radiation safety action should be justified by the value of the incremental collective effective dose avoided.' The principle of maintaining radiation dose ALARA has been introduced into radiation safety programs because of the prudent assumption that potential deleterious effects might occur a t any level of exposure, while recognizing that as the doses become smaller and smaller, the likelihood of a deleterious effect becomes vanishingly small. The concept of ALARA allows accounting for "social and economic factors" in determining an acceptable level of societal detriment for an activity. It is a principle by which the collective effective dose, and presumed detriment associated with an activity, may be constrained. Although individual doses should be controlled below the dose limits, there is no specific or unique value of dose for a task or occupational category that can be defined as "ALARA,"and the principle of ALARA is not a quantitative standard of care for individual workers or individual members of the public. "I'he costs related to an adequate design that complies with all current building codes and architectural standards are not associated with the application of the ALARA principle.

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2. APPLICATION OF ALARA

2.2 Concepts of a Cost-Benefit Approach to ALARA Three basic concepts that affect the productive application of the cost-benefit approach to the ALARA principle are: 1. use of collective effective dose (person-Sv) as a quantitative measure of objective health detriment 2. the magnitudes and distribution of individual doses that contribute to a specific collective effective dose value 3. the monetary value of the dose avoided 2.2.1

Applicability of Collective Effective Dose

The collective effective dose is the appropriate radiation quantity to be used for most risk assessments; however, there are practical limitations to its application. Estimation of collective effective dose requires definition of the sizes of various age and sex groups and of the pathways by which they are exposed (NCRP, 1995b). Collective effective dose should be used for risk assessment with caution if both the exposed population and the radiation doses can not be well characterized. Definition of the exposed groups and their modes of exposure is relatively straightforward in the occupational setting. Application of collective effective dose in the environmental arena is more challenging. It may not be feasible to define the collective effective dose with confidence if projection of population sizes and locations is required for times that are more than a few decades in the future. To determine the reasonableness of such assessments, uncertainties in both demography and in dosimetry must be identified and carried through the calculations to estimate the overall uncertainty in collective effective dose. If the relative uncertainty in collective effective dose is more than an order of magnitude, the estimate of collective effective dose is not adequate for making decisions (NCRP, 1995b). When the uncertainty for a projected potential collective effective dose is very large, it may be more appropriate to estimate risks to typical individuals in a critical group of people who might be exposed in the future. 2.2.2

Dose Magnitude and Distributions

The concept of a n individual dose that is negligible because it implies a n individual risk that can be ignored in the context of

2.2 CONCEPTS O F A COST-BENEFIT APPROACH TO ALARA

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everyday life has been defined by the NCRP. The value of the negligible individual dose is taken to be 0.01 mSv annual effective dose per source or practice (NCRP, 1993b). However, for collective effective dose calculations, all doses should be included, no matter how small because the use of the no-threshold dose response model logically implies that all doses contribute to the total risk (NCRP, 1995b). Examination of the distribution of doses that contribute to the collective effective dose is a n important step in any assessment. If the distribution is very broad, separation of the distribution into reasonably sized groups of persons with smaller ranges of doses is advisable. The collective effective dose may be dominated by exposures to one or more groups, while doses to other groups may be very small. It is appropriate to focus attention and resources on dose reduction for groups receiving the largest doses. As discussed in Section 2.2.3, when doses to some groups approach dose limits, the upper end of the distribution of doses should receive more attention in ALARA evaluations. Section 2.3 addresses the issue of the level of effort that should be devoted to ALARA evaluations of small collective doses. 2.2.3

Monetary Value of Auoded Dose

The ICRP (1983; 1989) recognized that the potential detriment caused by radiation exposure consists of a t least two components. The first component is the "objective health detriment," including all stochastic health effects for which quantitative estimates of the probability of occurrence as a function of radiation dose have been derived from exposed populations. These effects are primarily fatal and nonfatal cancers and birth defects. For purposes of radiation safety management, the dose-response function for the "objective health detriment" is assumed to be linearly proportional to collective effective dose and without a threshold. For this portion of the collective detriment, the value of the detriment per unit dose is a constant (a). The second component of detriment includes social factors and possible health detriments that reflect such factors as anxiety over individual levels of dose, uneven distribution of doses, the perceived risks of the doses, and concern on the part of management when individual doses are significant fractions of authorized limits (ICRP, 1983). For this portion of the collective detriment (P), the value of the detriment may be a function of dose and therefore may

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2. APPLICATION OF ALARA

change over the range of doses included in the collective effective dose assessment. ICRP (1989) defined the total detriment resulting from the use of radiation by a practice, a t an installation or from a specific radiation source as:

Y = aS+CbSj

(2.1)

j

where: a = the monetary value of the objective health detriment per unit of collective dose S = the total collective effective dose Sj = the collective effective dose originating from a per caput dose Hj delivered to the Njindividuals of the jth group Pj = the value of the collective detriment assigned to a unit of collective effective dose in the jth group Unless doses approach either legal limits or internally imposed constraints, the second component of detriment may be very small in comparison with the value of the objective health detriment and can usually be ignored. In that case, the ALARA principle would lead to implementation of an action that would reduce the colleca t a cost not exceeding the tive effective dose by an increment (AS) quantity (M). When individual doses are near the limit appropriate for the exposed population, considerations other than the objective health detriment may justify additional expenditures for dose reduction. Justification for these choices will vary from one organization to another and may depend upon assessments of parameters that are specific to a particular practice or industry. Although the NCRP does not recommend nor endorse any specific values for a or pj, the examples used by the ICRP (1989) illustrate the numerical application of these concepts. For all dose ranges, the value of a is assumed to be $20,000 (person-~v)-l.Values of pj were defined for three individual dose ranges: 1. For groups with individual doses of <5 mSv, PI = $0 (person-~v)-l 2. For groups with individual doses in the range 5 to 15 mSv, Pa = $40,000 (person-~v)-l 3. For groups with individual doses in the range >15 to 50 mSv, PB = $80,000 (person-~v)-l

2.3 SCREENING FOR hLARA ASSESSMENT

/ 11

2.3 Screening for ALARA Assessment

When the ALARA principle is applied, the cost of the assessment of risk should be included in the optimization. The effort expended in assessing the risk should not be disproportionate to the risk itself. An obvious threshold for optimization occurs when collective effective dose is so small that the benefit obtained from its complete elimination would not justify the cost of evaluation. A simple mechanism should be used to determine whether the potential collective effective dose related to a proposed practice, procedure or situation is likely to exceed this conceptual threshold. Direct measurements of exposure rates (or of concentrations of radioactivity in air) are appropriate as screening measurements to determine if an evaluation of the application of ALARA is needed. A screening level for a minimal level of documentation of the application of the ALARA principle for occupational exposures can be estimated. While the value of some dose reduction actions may be apparent from a simple mental calculation, an avoided collective effective dose of the order of 0.01 person-Sv appears necessary to justify an optimization evaluation that entails formal procedures. This estimate assumes that the doses are reasonably distributed among individuals and that none of the occupational doses approaches a limit. Additionally, the formal procedures and documentation needed to implement ALARA should also be minimal if the expected collective effective dose lies below 0.01 person-Sv. For collective doses of less than 0.01 person-Sv, the total value of the dose that might be partially avoided by a formal A U R A program does not justify the effort required for the preparation of formal procedures and documentation. However, less formal efforts to maintain doses below that level may still be justified. For a practice that results in exposure of the general public, similar considerations apply. A determination of whether projected doses to individuals approach appropriate limits or are very unevenly distributed is a first step. However, a study of alternatives that could reduce dose to the public may well be more complex than a n evaluation of a workplace improvement.

3. Organization and Administration 3.1 Management Commitment and Policy

The highest level of management of a n organization is responsible for establishing the goals of the organization and for ensuring that it has sufficient resources to safely cany out the operations needed to meet these goals. It also has the responsibility to ensure that an effective radiation safety policy is established and that the radiation safety program is implemented for the protection of employees and the public. All employees who may be occupationally exposed should be informed of the radiation safety policy and programs. It is especially important that all employees be thoroughly instructed in and understand their responsibilities in support of the radiation safety policy and program. The radiation safety policy should define the goals of the radiation safety program, including the organization and administrative control required for the use and handling of radioactive sources and radiation-producing equipment. It should state a commitment to the application of the ALARA principle and the adequate and cost-effective control of radiation exposure of workers and the public. This policy should represent a commitment by management to provide suitable budgetary support for the radiation safety program. Effective implementation of the policy requires that good radiation safety practices be followed and that the applicable regulatory requirements2 be met. Activities that involve radiation or '1n the United States, federal regulatory requirements pertaining to radiation safety have been established by the Nuclear Regulatory Commission (NRC), Department of Energy (DOE), Environmental Protection Agency (EPA), Department of Transportation (DOT), Food and Drug Administration (FDA), Occupational Safety and Health Administration (OSHA), and the Federal Emergency Management Agency (FEMA). State agencies have varying degrees of additional regulatory requirements, and, in some cases, act on behalf of a federal agency.

3.2 RADIATION SAFETY ORGANIZATION /

13

radioactive materials should be under periodic surveillance by work area supervisors, the radiation safety staff, and safety auditors. Operational records should reflect safety-related actions (NCRP, 1992). Management support of worker training is essential. NCRP Report No. 105 (NCRP, 1989a), for example, states that an effective training program requires significant involvement by everyone, management and workers, not just the radiation safety staff. Guidance is given for medical and dental activities in Reports Nos. 105 and 107 (NCRP, 1989a; 1990) which can be adapted for other radiation programs as well.

3.2 Radiation Safety Organization The key to an effective program is the formal delegation of authority to competent staff members. The manager of the radiation safety program may be referred to as a Radiological or Radiation Safety Officer (RSO), Radiological ControI Manager, Radiation Protection Manager, or some other title. For this Report, this manager will be referred to as the RSO. The RSO should be directly responsible to the highest level of management and should have ready access to all levels of the organization.

3.2.1

Radiation Safety Advisory Organization

Management should appoint a Radiation Safety Advisory Group. Certain regulations may specify the establishment, membership and responsibilities of such a group. For this Report, the group is referred to as the Radiation Safety Committee (RSC). The responsibi1it;y of the RSC is to formulate institutional radiation safety policies, review and audit the effectiveness of the radiation safety program, and provide guidance to the RSO on the operational uses of radiation and radioactive materials. The RSC should include individuals who are knowledgeable about the use of radioactive materials and radiation-producing equipment in the facility. It may also include persons who are knowledgeable about the overall organization and its legal, financial, procurement and other business functions. The RSO should be a n ex offGcw member of the RSC. The RSC should perform reviews of the purpose, safety and compliance with the radiation safety program and regulatory requirements of all proposed work with radioactive material or

14 /

3. ORGANIZATION AND ADMINISTRATION

radiation-producing devices. Members of the RSC should be qualified in their normal field of endeavor. Management is responsible for assuring that members, and especially the chairman, have the necessary experience and qualifications to meet the responsibilities of the RSC.

3.2.2

Radiation Safety Oficer

The RSO is responsible for advising management concerning radiation safety practices and regulations. This individual should be delegated the authority to supervise the operational radiation safety organization, develop a budget and commit expenditures that are allowed by that budget. The RSO should have adequate funding to maintain a stable radiation safety program and should have access to consult the highest level of management on any concern regarding the radiation safety program. The RSO is responsible for periodic and special surveillance of activities such as acquiring and disposing of radioactive materials, training in radiation safety practices for facility employees and users, developing and maintaining radiation control and dosimetry records, and authorizing the use of radiation and radioactive materials within the facility. The RSO is also responsible for developing and maintaining a radiation safety manual. The activities of the RSO and the radiation safety staff should in no way remove or reduce the responsibilities of radiation users or line supervisors to conduct their work in a safe manner. The principal functions of the RSO and staff are training and support through the provision of common radiation safety services, including radiation safety surveillance. 3.3 Accreditation and Certification

The minimum qualification of the RSO will depend on the magnitude of the potential hazards and complexity of the operation. The RSO should possess a n appropriate academic background together with practical radiation safety experience germane to the operation. Specialized education in health physics a t the college level, combined with practical experience, is preferable. The American Board of Health Physics and the American Board of Medical Physics certify professional health physicists who meet their requirements. An individual who is certified or has equivalent qualification is generally considered qualified to serve as an

3.4 RADIATION SAFETY PROGRAM POLICIES AND PROCEDURES

/ 15

RSO for an organization utilizing complex and varied sources of radiation. The RSO is responsible for developing a qualified staff of radiation safety professionals and technicians of a size and level of expertise appropriate to the activities of the program. An effective on-the-job training program is essential for all employees. Continuing education programs relying on expertise within the organization as well as opportunities provided by professional societies, universities and other organizations should be used to maintain high skill and knowledge levels. Management should support and encourage staff members to become certified by appropriate organizations such as the American Board of Health Physics, the American Board of Medical Physics, the American Board of Radiology, or the National Registry of Radiation Protection Technologists. 3.4 Radiation Safety Program Policies and Procedures Radiation safety policies and procedures should be clearly stated in a radiation safety manual. Specific procedures should be written by operations groups and the radiation safety staff to implement the radiation safety policies for various tasks that are performed in the course of operating the facility. It is important to distinguish operating procedures employed by the radiation users and the radiation safety staff from the policies and procedures that are promulgated in the overall facility radiation safety manual. 3.4.1

Radiation Safety Manual

The radiation safety manual should include a comprehensive statement of policy and the principal administrative and program procedures established by the RSC. Significant technical support of the RSO is needed for writing this manual. The radiation safety manual is provided to help workers use radiation and radioactive material safely in compliance with the organization's policies and in compliance with regulatory requirements. The radiation safety manual should include: 1. management's commitment to proper radiation safety practice

16 /

3. ORGANIZATION AND ADMINISTRATION

2, description of the RSC, the radiation safety staff, and the radiation safety program 3. specific policy and regulatory requirements 4. specific procedures on how to comply with these requirements

Management should support periodic revisions of the manual including appropriate contributions from the RSC and the RSO. 3.4.2

Radiation Safety Operating Procedures

Radiation safety operating procedures that govern the radiation safety staff are prepared by the RSO. Lead managers or supervisors with the assistance of the radiation safety staff prepare specific operating procedures that govern activities and processes that involve the use of radiation or radioactive materials. Procedures are instructions that describe actions and steps necessary to conduct a particular task, actions and protective measures to conduct the task safely, and steps necessary to document performance of the task. Procedures provide a means for enabling consistent, reproducible work and may range from relatively simple instructions to complex manuals. Within the radiation safety staff, for example, there should be procedures covering instrument calibration and use, laboratory sample counting, radioactive waste handling, and the calibration and use of personnel dosimeters as appropriate. Depending on the complexity of a particular task and the training and experience of the individuals involved, procedures for work that involves radiation or radioactive materials should include the following elements as appropriate:

i

?

1. a description of the work t a t is authorized 2. a description of the potent a1 hazards that will be encountered in performing the work, including potential radiation dose rates, identification of the sources of radioactive material, potential radioattive contamination levels, and the potential for intake of nadioactive material 3. the identification of indididuals responsible for making sure that the work activitiks are conducted in accordance with the safety procedure 4. the safety controls and procedural safeguards that are necessary to prevent or limit exposure including requirements for protective clothing

3.5 RESPONSIBILITY /

5. 6.

7. 8. 9. 10. 11.

17

respiratory protection internal and external dosimetry radiation surveys worker time and dose limitations limiting conditions for either radiation or contamination levels health physics or radiation safety coverage that is required during the task required worker qualifications including any specialized training actions to be followed in the event of an emergency a description of contamination control requirements a description of required training and tasks that should be completed before beginning the task a t hand a description of the method for authorizing deviations from the specified procedure references to records and reports to be completed a description of acceptable results and of actions to be taken in response to unsatisfactory results

Procedures can be simple and brief, but those that govern the use of high activity radiation sources or the use of large quantities of radioactive material are usually detailed and complex. Procedures should always be based on established policies, good practices, and regulatory requirements. Sources of information helpful in developing procedures include the recommendations of the NCRP, the ICRP, the International Atomic Energy Agency (IAEA); the consensus standards of the ANSI, the American Society for Testing and Materials (ASTM), the American Conference of Governmental Industrial Hygienists (ACGIH), and the American Industrial Hygiene Association (AIHA); the suggestions from NRC regulatory guides and the guides of the Conference of Radiation Control Program Directors; and other guidance documents. 3.5 Responsibility

Ultimately, workers are responsible for their own safety. However, supervisors and managers are responsible for providing a safe workplace and for promoting an attitude of responsibility for safety among all workers. Compliance with radiation safety procedures and demonstration of individual responsibilities described in the overall program policy should be a n established performance expectation for every individual involved, both workers and man-

18 / 3.ORGANIZATION AND ADMINISTRATION

agement. The RSC and RSO are accountable to management for developing and implementing a radiation safety program that meets the radiation safety policy and regulatory requirements, and that supports the implementation of the ALARA principle. Management is responsible for establishing and funding a radiation safety organization that includes independent quality assurance checks and that is adequate for the complexity of the operations utilizing radiation and radioactive material. 3.6 Quality Assurance

Management should ensure that there is a quality assurance program in place to provide oversight of the radiation safety program. A quality assurance program should encourage and support self-assessments within the work group to ensure that the radiation safety goals are met. Independent audits should be developed when necessary to gain an outside perspective of any aspect of the radiation safety program. The overall goal of a quality assurance audit program should be improvement of performance. The program should never be an adversarial or fault-finding activity. Quality assurance is a systematic evaluation of activity to measure outcomes against expectations. Audits, inspections, surveillance and statistical evaluations (quality-control checks) are all basic quality assurance tools used to make systematic evaluations. A radiation safety program should contain quality assurance assessment to evaluate the adequacy of:

1, basic control of radiation-producing equipment and radioactive materials 2. conformance with organizational policies and regulatory requirements 3. contamination and effluent control, and protective measures 4. health physics assessment in the workplace and surrounding environment 5. health physics assessment of dose to workers and the public 6. incident and accident investigation and corrective actions 7. training 8, record keeping

3.6 QUALITY ASSURANCE / 19

3.6.1

Management Goals

Management has a responsibility not only to assure that corporate goals are met, but also to assure that operations are performed with safety, reliability and cost effectiveness. The assurance of safe, reliable and cost-effective performance depends on management, including the RSO. Managers and supervisors have the responsibility to provide a safe work environment and to ensure good work practices in the areas that they manage. Reviews of operating records and personal interviews with the staff are necessary components in the discharge of this responsibility.

3.6.2

Surveillance

Radiation safety surveillance is a quality assurance activity, although it may also be necessary to demonstrate regulatory compliance and increase worker and public confidence. It depends on training workers to establish a high quality of performance in proper handling, controlling and monitoring of radiation and radioactive material. Area surveys and personal monitoring are significant aids for determining the adequacy of facility design, operating procedures, and worker training. A high-quality surveillance program depends on the availability of functioning and calibrated instrumentation (see Section 10). The RSO should expect prompt, accurate and consistent reports of the results of routine area surveys and personal monitoring. These reports can provide a n indication of serious inadequacies in the facility procedures and training. Records of off-standard conditions should be written clearly on the survey form or in the survey log book, and confirmed by the responsible health physicist. Any corrective action that is recommended and implemented should be documented and considered to be provisional until reviewed by a health physicist or the RSO. Routine surveys and personal monitoring are usually done on a regular schedule, but may be relatively infrequent (weekly, monthly or quarterly). For this reason, it is important that supervisors understand their essential role in controlling radiation exposure and in recognizing the implications of changes in operating conditions. This is especially critical when high-dose rate radiation sources are being used. Supervisors and workers should be well trained in the detection and control of radioactive contamination when it is a possibility in the workplace. Follow-up and special surveys should be performed to:

20 /

3. ORGANIZATlON AND

ADMINISTRATION

1. confirm the findings of a routine survey that identified off-standard conditions 2. assist in the investigation of individual exposures that are greater than expected 3. confirm unexpected findings of radioactive contamination, assess any mishandling of radioactive waste, or locate missing radioactive material 4. assure success of decontamination efforts 5. verify adequacy of shielding 6. verify measurements made during regulatory inspections 7. provide training for new radiation safety staff

In addition to the above, special surveys may be required as part of an investigation of a major radiation accident or as a result of concern on the part of any worker. These investigations require the active participation of the radiation safety staff. A thorough investigation report should be prepared and is usually required by the responsible regulatory authority. 3.6.3

Program Audits

A radiation safety program audit is a deliberate examination of the program to determine if it is effective. The audit is an integral part of any quality assurance effort and should not be confused with the entire radiation safety effort. Two common types of audits are those that are done by the RSO and persons within the radiation safety program (self audit), and those that are done by persons from outside the program (independent audit). The major advantage of the self audit is that problems can be identified and promptly corrected by the workers, supervisors and radiation safety staff. The self audit should be an activity performed by a team selected from the workforce, the RSC, supervisory health physicists, and the RSO who have the knowledge and experience to identify and correct any problems or deficiencies. Self audits should be scheduled to provide a systematic annual review of the entire radiation safety program. Problems identified and defined from self audits should be corrected in a timely manner and lessons learned should be examined for applicability to other parts of the radiation safety program. An independent audit should be scheduled as needed to assess the overall program. Because this audit has the advantage of a n independent perspective, deeply entrenched work methods that may be overlooked during self audits can be identified. Further, in

3.6 QUALITY ASSURANCE / 2 1

auditing measurement procedures and results, the independent auditor may suggest improvements or changes in standards and techniques. The independent audit should be preceded by adequate notification, which includes the time and purposes of the audit, the work areas to be covered, the persons to be interviewed, and the records needed for review. Ideally, an independent audit should find no deficiencies that have not been identified by self audits. Both types of audits require preparation, observation, evaluation and communication. Audit reports should be factually correct and should emphasize the relative importance of the various findings. Fault-finding, blaming and adversarial relations should be avoided. The overall goals of an audit are to ensure that the intended work quality is being maintained and that possible improvements are identified and fostered. 3.6.4

Incident and Accident Investigations

The investigation of incidents and accidents must be timely. Both internal reports for management and reports to regulatory bodies may be needed in preliminary form within hours or days and as final reports within days or weeks. These reports are the responsibility of the RSO and the supervisor who is responsible for the operation that was involved in the incident or accident. Incident and accident investigations should include a thorough examination of the scene, interviews with the people involved, a review of pertinent records, and a complete and accurate report of the incident or accident and subsequent investigation. The location of the event should be completely surveyed with appropriate instruments as needed to determine and document the radiation levels and the extent of radioactive contamination. Personal monitoring devices should be collected and evaluated, and bioassays should be performed as needed. An inventory of all radioactive material and waste should be made. Any records or logs that have been maintained should be examined. Workers and others in the area should be interviewed early in the investigation. A photographic record of the area may be important to reconstruct the incident or accident, e.g., in the case of a serious individual exposure. Although incident and accident investigations are important both legally and for program improvement, avoiding unnecessary interruptions of important activities (e.g., medical service, production and research) is also important. However, in some cases the RSO may decide to suspend work in an area to allow an immediate

22 / 3. ORGANIZATIONAND ADMINISTRATION

and thorough investigation. Expeditious completion of the investigation would permit an early resumption of work. When important activities are interrupted, appropriate notations in the operating records are necessary. The records should always be sufficiently detailed to allow future review. It is also important to remember that incident and accident reports could be made available for public scrutiny. 3.6.5

Deficiency Dacking

Improvements and corrections must be made occasionally in all radiation safety programs because of changes in: 1. organizational mission

2. regulatory or contractual requirements for which the institution is responsible

3. work objectives, priorities or efficiency 4. technical knowledge or support

5. personnel 6. deficiencies discovered during operations, surveys, audits or facility reviews Corrections of deficiencies and improvements in operations often arise from reviews of the program by the RSO and from self audits. Some may arise from the independent audits or the recommendations of government regulators. In any case, changes in the program and the persons responsible for effecting the changes must be recorded. This is especially important to ensure that there is a clear understanding of responsibility, a reasonably uniform approach to safety, and a well-documented response to the identified deficiency. Changes or corrective actions must be communicated to the RSC and to appropriate employees. This is especially important when common services such as personal monitoring, waste management, training, and emergency response are affected. Records help to limit the confusion that may arise during independent audits and ease the quality assurance efforts of the RSO. Good records are important for effective tracking of changes by the RSO and management.

3.8 OCCUPATIONAL MEDICINE /

23

3.7 R e c o r d s M a n a g e m e n t The amount and detail of the records that the RSO should maintain has become substantial and their maintenance represents a n appreciable portion of the effort of the radiation safety staff. The main records are "master" copies of the radiation safety policy manual and the radiation safety procedures. Included in the records that should be maintained are those that detail administrative actions that affect the program, report internal and external audits, and record deficiencies and corrective actions. Operating procedures, personal monitoring and survey records, instrument calibration records, waste management records, and records of worker training should be maintained in a readily retievable form. The NCRP has provided guidance for maintaining radiation safety records in Report No. 114 (NCRP, 1992).

3.8 O c c u p a t i o n a l M e d i c i n e The health of the worker is essential to the effective functioning of the organization. Therefore, a good occupational health program should be provided. Special health evaluations and monitoring may be needed for occupationally exposed individuals working in special environments, e.g., high temperature areas or high airborne contamination areas. Special health services should be available to provide care to workers accidentally exposed to high radiation doses or to high internal or skin contamination. While no special health care is needed for workers who receive normal occupational exposure, concerns about radiation exposure may occur under special circumstances, e.g., assignment to an unfamiliar job, a necessity to receive more radiation exposure than usual, exposure to unfamiliar conditions that involve sources of radiation and pregnancy. For this reason there should be persons available who are knowledgeable about the risks of radiation exposure and who are trained to address these concerns. Management and the staff of the radiation safety program should be sensitive to the concerns of the workers.

24 /

3.

ORGANIZATION AND ADMINISTRATION

3.9 Recommended Additional Reading

A Compendium of Major US.Radiation Protection Standards and Guidelines: Legal and Technical Facts, ORAU 88/F-111, Oak Ridge Associated Universities, Washington, 1988. Dose Control at Nuclear Power Plants, NCRP Report NO. 120, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1994. The Handling, Storage, Use and Disposal of Unsealed Radionuclides i n Hospitals and Medical Research Establishments, International Commission on Radiological Protection Publication 25, Annals of the ICRP 1, Pergamon Press, Elmsford, New York, 1977. Limitation of Exposure to Ionizing Radiation, NCRP Report No. 116, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1993. Management of Radioactive Material Safety Programs at Medical Facilities, Draft Report for Comment, Camper, L.W.,Schlueter, J.,Henderson, P., Bermudez, H., Fuller, M., Jones. J.,Campbell, V., Montgomery, J. and Allen, K., NUREG-1516, U.S. Nuclear Regulatory Commission, Washington, 1995. Occupational Dose Reduction at Department of Energy Contractor Facilities: Bibliography of Selected Readings in Radiation Protection and ALABA 5, DOEIEH-0364T, BNL-43228, U.S. Department of Energy, Washington, 1994. Occupational Dose Reduction at Department of Energy Contractor Facilities: Study of A U R A Programs - Good Practice Documents, DOEIEH-0278T, BNL-47339, U.S. Department of Energy, Washington, 1992. Occupational Dose Reduction at Nuclear Power Plants: Annotated Bibliography of Selected Readings i n Radiation Protection and ALARA 8, NUREGICR-3469, BNL-NUREG-51708, U.S. Nuclear Regulatory Commission, Washington, 1996. Operational Radiation Protection: A Guide to Optimization, IAEA Safety Series No. 101, International Atomic Energy Agency, Vienna, 1990. Operational Radiation Safety-Training, NCRP Report No. 71, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1983. Optimization and Decision-Making i n Radiological Protection, International Commission on Radiological Protection Publication 55, Annals of the ICRP 20, Pergamon Press, Elmsford, New York, 1988.

3.9 RECOMMENDED ADDITIONAL READING / 25

Provision of Operational Radiation Protection Services at Nuclear Power Plants, LAEA Safety Series No. 103, International Atomic Energy Agency, Vienna, 1990. Quality Assurance for Diagnostic Imaging, NCRP Report No. 99, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1988. Radiation Protection i n Occupational Health, IAEA Safety Series No. 83, International Atomic Energy Agency, Vienna, 1987. Radiation Protection i n the Mineral Extraction Industry, NCRP Report No. 118, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1993. Radiation Safety Training Criteria for Industrial Radiography, NCRP Report No. 61, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1978. Radiological Control Manual, DOE5H-0256T, U.S. Department of Energy, Washington, 1992. Radiological Protection i n Biomedical Research, International Commission on Radiological Protection PubIication 62, Annals of the ICRP 22, Pergamon Press, Elmsford, New York, 1991. Ra'diological Protection of the Worker i n Medicine and Dentistry, International Commission on Radiological Protection Publication 57, Annals of the ICRP 20, Pergamon Press, Elmsford, New York, 1989. Recommendations for the Safe Use and Regulation of Radiation Sources i n Industry, Medicine, Research and Teaching, IAEA Safety Series No. 102, International Atomic Energy Agency, Vienna, 1990. Suggested State Regulations for Control of Radiation, Volume 1, Ionizing Radiation, Conference of Radiation Control Program Directors, Frankfort, Kentucky, 1997.

4.

Facility Design

Properly designed facilities allow for a much higher degree of safety than can be obtained by dependence on administrative rules and procedures in inadequate facilities. While good design can never eliminate the possibility of accidental radiation exposure or contamination, the probability and magnitude of such occurrences can be greatly reduced. Proper facility design is also the most effective approach in reducing unnecessary occupational exposures or releases to the environment. Attention to the radiation safety and control aspects of facility design can minimize later operating difficulties. The planning and design of new or modified facilities should include review by qualified experts to ensure that appropriate radiation safety features are incorporated. Competent input and review in these stages will facilitate operation within established safety standards and maintain radiation exposure a t levels that are ALARA with minimal adverse operational effects. 4.1 Site Selection

There are several factors that should be considered in the site selection process for a facility designed to handle radioactive material or to use radiation-producing equipment. Many of these factors are based on radiation safety considerations and relate to the radiation doses that could be received by workers or members of the public outside the facility during routine operations or accidental conditions. A facility for handling radioactive material or using radiation-producing equipment must be located and designed so that the radiation doses to persons outside the facility can be maintained below applicable limits and are ALARA. While proper design can preclude or minimize the release of radioactive materials and the emission of direct radiation from the facility, the site itself can provide a n additional margin of safety. The potential risk posed by activities within a facility to persons outside the facility will dictate

4.1 SITE SELECTION

/ 27

the importance of site selection. The greater the potential risk, the greater the importance of site selection. For facilities where radioactive material will be used, the specific factors to be considered in site selection include: 1. the types and quantities of radioactive material to be handled 2. the physical and chemical forms of radioactive material 3. the type of work to be performed 4. the potential for dispersal of radioactive material to the environment during routine operations or accidental conditions 5. the potential pathways for release of radioactive material (airborne, liquid, contaminated solids) 6 . the levels of direct radiation that could exist outside of the facility during routine operations or accident conditions

For facilities where radiation-producing equipment will be operated, the specific factors to be considered in site selection include: 1. the types of radiation to be produced 2. the energy and intensity of the radiation 3. the levels of direct radiation that could exist outside of the facility during routine operations or accident condtions 4. the potential for the production of activation products 5. the potential for the release of activated material to the environment during routine operations or accident conditions 6 . the potential pathways for release of the activated material (airborne, liquid, contaminated solids) It is not uncommon for facilities both to handle radioactive material and to use radiation-producing equipment. In such cases, all of the above factors need to be considered in site selection. If the quantities of radioactive materials that could be released from either type of facility under normal or accident conditions could result in exposure a t significant fractions of dose limits to individuals outside the facility, the meteorological, hydrological and seismological characteristics of the site must be evaluated. These factors always have to be considered for large facilities like power reactors, and radioactive waste processing facilities. Also common to both types of facilities is the need to consider the spatial and temporal distribution of potentially exposed individuals. That

28 /

4. FACILITY DESIGN

is, the density of the relevant population as well as the associated occupancy factors should be considered. Other factors that need to be considered in the site selection process include economic considerations, public or commercial access, the risk from hazards related to chemicals, explosives, carcinogens and other such materials, zoning and other regulatory requirements, and public acceptance. For example, it may be economical to locate a radiation oncology suite in the basement of a building, since i t is likely to need less additional shielding than one located a t ground level. Similarly, a large high energy accelerator located in a n isolated area might require less shielding to limit doses to members of the public than one located on a crowded campus, because a greater distance could be maintained between the source and potentially exposed individuals.

4.2 Facility Layout Facility layout is a n important aspect of design and an inherent aspect of the implementation of the ALARA principle. Radiation safety must be recognized a s an integral part of any operation that uses radiation sources or radioactive materials, and space for this function should be allocated within the facility, as appropriate. Areas designated specifically for radiation safety functions, e.g., storage of monitoring instruments and supplies, analyzing and recording of surveillance samples (air filters, wipes, etc.), instrument calibration, decontamination of workers or equipment, processing and packaging of radioactive wastes, etc., should be integrated into the overall facility layout. Functional portions of the facility need to be located properly, relative to each other - for efficient operations, for ease of movement of people and materials into and out of processing areas, and for effective maintenance. When practical, the facility should be designed with zones based on gradients of potential external radiation exposure, potential airborne or surface contamination, and similar considerations. For example, lunch rooms, offices and conference rooms should be located in clean zones. Zones with minimal risk of exposure include sample counting rooms and low-level radioisotope laboratories. Laboratory or processing areas containing the largest quantities of radioactive materials should be the most isolated from the clean areas. High-intensity radiation sources should be surrounded by low-occupancy areas to reduce the risk of inadvertent exposures. Persons going from one clean area to

4.3 EQUIPMENT AND SYSTEM DESIGN /

29

another, or between low-risk areas, should not be required to pass through zones of greater risk of exposure or contamination. Protective clothing change rooms and areas for personal contamination monitoring and decontamination should be established at points of access to areas used for handling significant quantities of radioactive materials. Attention must also be given to control of potentially contaminated persons in the event of an accident requiring rapid egress from the facility. At least one area for personal contamination monitoring and decontamination should be accessible without requiring individuals to pass through an area that may have become highly contaminated as the result of an accident. To the maximum extent compatible with facility operations, uses of radioactive materials should be segregated from uses of hazardous chemicals. Waste management and disposal of mixed wastes is much more difficult than dealing with either radioactive or hazardous chemical wastes separately. 4.3 Equipment and System Design

In addition to the facility layout, it is important to design or select specific f ~ t u r e s equipment, , systems and components for effectiveness in relation to radiation safety and operational efficiency. Equipment and components that may become highly radioactive or contaminated should be designed for accessibility, ease of maintenance, ease of installation and removal, and ease of decontamination (ASTM, 1991). The provision of fixtures designed to support removable shielding or containment enclosures can reduce the time spent in the vicinity of activated or contaminated equipment. At accelerator or reactor facilities, it is prudent to choose materials that will not become highly radioactive for use in areas where exposure to particle beams or neutron radiation is likely. A decommissioning plan should be developed in the facility design stage to guide the selection of materials for the facility and equipment. The operations planned for the facility should guide the selection of installed monitoring and surveillance equipment. Installed radiation monitoring systems are generally necessary in facilities in which operations can cause the radiation levels in potentially occupied areas to reach dose rates that would be of concern. They can also be useful for remote monitoring of highly radioactive filters, ion exchange resins, tanks and other equipment. These instruments must be carefully chosen based on the radiation fields and

30 / 4.FACILITY DESIGN

dose rates that they will be required to monitor. For example, while Geiger-Miiller detectors would be appropriate for radiation fields produced by sources emitting gamma rays, ionization chamber detectors would be needed for pulsed x-radiation fields. In some facilities, neutron detectors may be required. Installed monitors may be needed in some facilities a t the exits from areas in which loose radioactive material is handled. These monitors can be designed to detect radioactive contamination on hands and feet, or they can check the entire body as the individual leaves the area. In some cases installed monitors may also be required for checking equipment that is being removed from an area in which it could have become contaminated or activated. These monitors may also be useful for detecting radioactive material that may have been put into normal trash. If the dispersal of radioactive material is possible, installed monitors for detecting airborne radioactive material may be needed. At facilities in which extremely high or potentially lethal doses of radiation can occur, an access control and warning system is necessary. Such a system should be designed to address the hazard that is present. These systems can be quite simple, such as warning signs, a locked door, or a radiation detector alarm. They can also be very complex key-controlled or computer-based programmed logic interlock systems. Recommendations on the specification and design of these systems are contained in NCRP Report No. 88 (NCRP, 1986). 4.4 Shielding

This Section is not intended to be a manual for shielding design. However, some of the important considerations are briefly discussed. Shielding may be necessary to reduce the potential for exposures to workers and visitors a t the facility and to the public in the vicinity of the facility. Because individuals may be exposed to multiple sources of radiation, it is recommended that shielding for any single source be designed to limit radiation exposure to some fraction of the recommended dose limit (NCRP, 1993b). Designers should recognize the uncertainties inherent in estimating the potential radiation levels and in calculating the transmission of radiation through the shielding material. These uncertainties are larger when the radiation field is complex, e.g., where there are mixed neutrons and gamma rays or where there is a wide spectrum of energies involved. Composite shields, i.e., shields made of

4.4 SHIELDING

1 31

mixtures of materials, can also lead to significant uncertainties in estimating radiation transmission. Both the cost and spatial extent of shielding should be considered, but radiation safety must not be compromised. Placing shielding close to the source will reduce the volume of material needed. Often the amount of shielding can be reduced by installing engineered controls that restrict access to areas immediately outside the shield wall, or areas above or below the facility. Consideration should also be given to the amount of time that the source is exposed o r the radiation-producing equipment is on. The possibility of increased use of the source or future design changes that would increase access near the source should be taken into account. The field of shielding design includes considerations of radiation physics, materials properties, and structural engineering. Important references dealing with shielding include: Blizard (1962), Blizard and Chilton (1968; 1970), Burchsted et al. (1976), Chilton et al. (19841, Goldstein (1959), Jaeger (1975), McGinley (1993), Price et al. (1957), Profio (1970), Rockwell (1956), and Shultis and Faw (1996). Data necessary for shielding design and recommendations concerning shielding techniques are also contained in several NCRP reports (NCRP, 1976a; 1977; 1978a; 1983b) and ANSI standards (ANSWANS, 1985; 1991; 1997). In summary, various materials can be used for shielding, depending on the type of radiation, its energy and intensity, and the attenuation required. Typically medium and high atomic number materials such as iron and lead are effective for shielding x and gamma rays. For moderating fast neutrons a material with a high hydrogen content, such as water or polyethylene, must be included in the design. When thermal neutrons are captured in hydrogen, cadmium or other elements, high-energy gamma rays are emitted and must be considered in the shield design. Concrete is suitable for shielding both photons and neutrons and is a cost-effective material of choice when space is available. Earth is also an effective and inexpensive material that is widely used as shielding in various types of facilities. In addition to meeting radiation protection goals, the selection of shielding material is dependent upon engineering factors such as weight, cost, structural stability and compatibility. Shielding design should be an integral part of the initial planning for the facility. Ultimate removal and disposal of shielding should also be considered during the design phase. For example, lead shielding may create a hazardous waste disposal problem upon removal. In some experimental facilities the need for

32 /

4. FACILITY DESIGN

flexibility will be a primary consideration. In these situations, the design must ensure that future additions and reconfigurations of the shielding can be accommodated. If this is not done, future changes may be difficult and may not achieve the desired reduction in dose rate without significant additional cost and space.

4.5 Ventilation

This Section is not intended as a manual for ventilation system design. However, some of the important considerations related to radiation safety are discussed. Extensive work has been done on the design and engineering of ventilation for hazard control. Some important references dealing with ventilation are: ACGIH (19981, ANSI (19791, ANSUAIHA (19921, ANSIJASHRAE (19891, Cooper and Alley (19941, Kathren et al. (1980), McDermott (19851, Plog (19881, and UL (1990). Proper ventilation is necessary to control the movement of airborne radioactive materials, and to prevent or minimize the spread of contamination within the facility. In addition, ventilation systems are designed to control concentrations of airborne radioactive materials in the work areas, thus reducing the probability of internal exposure. Facilities in which unsealed sources or loose contamination may be present should incorporate design features to prevent the buildup and spread of contamination and airborne radioactive materials. The facility ventilation system is of primary importance in the control of contamination and airborne radioactive materials. Operations that routinely produce airborne contamination should use engineered containment and ventilation systems to prevent exposures to individuals from airborne releases to the environment. When possible, facilities should be designed so that use of a personal respiratory protection device is not required for normal operations. Appropriate personal respiratory protective devices may be used in accordance with the requirements specified in Section 7.5.2, but only in abnormal situations or when effective engineering controls are not feasible. Respiratory protective devices are not an acceptable substitute for engineered ventilation systems. For radiation safety, the primary functions of a ventilation system are to move airborne contamination away from occupied work areas (and the potentially exposed workers) and to provide a mechanism for the "recontainment" of the airborne radioactive material that was released. To meet these objectives, the ventilation system must have acceptable pressure differentials between work areas

and the outside environment. High-efficiency particulate air (HEPA) filtration or other appropriate filtration (e.g., charcoal filters) may be needed, but the radiation exposure of individuals from the radioactive materials retained on the filter should be evaluated. A pressure differential system should be used to control the flow of airborne contamination. In the system design, a pressure gradient should be established, with the lowest pressure and collection points in areas with the highest potential for release of dispersible material. The flow should always be from clean areas to contaminated areas, but it must be recognized that similar areas may not always require the same ventilation characteristics (e.g., pressure differential and filtration), and the system design should provide some flexibility. The design of the ventilation system should provide for proper air flow under all conditions, including open and closed positions of doors and other openings and changes in system setup. Recirculation of air should be avoided unless the system has been specifically designed for such use and proper filtration and monitoring are provided. Air cleaning devices are often incorporated into the system design. Charcoal absorbers and chemical scrubbers may be used as appropriate to reduce the concentration of volatile gases and organic radiochemicals in effluent. A single-stage HEPA filter may be desirable in areas in which airborne radioactive materials are not normally expected but might be present during accidents. For facilities that contain unsealed highly radiotoxic material or radioactive materials in dispersible form, a multi-stage HEPA filter is recommended. Exhaust air filters may be needed to ensure that releases do not exceed applicable administrative control limits. When filtration systems are used, each filter stage should be designed and located to facilitate independent testing according to applicable standards. The design should include considerations of filter replacement. Proper design will allow the filters to be changed easily while minimizing the potential for release of radioactivity and worker exposure. Exhaust vents and stacks should be located carefully to avoid recirculation of exhaust air through ventilation system intakes. The design should also include provision for modifying the ventilation during a n accident, e.g., containment, use of a redundant system, use of a by-pass system, or change in flow rates. Controls for the ventilation system should be located in areas that are readily accessible in the event of an accident.

34 /

4. FACILITY DESIGN

To reduce the impact of the potential release of radioactive materials into a work area, it is common to use chemical fume hoods or gloved boxes as the first containment barrier for these materials. Design of a hood (or a gloved box) for radioactive material use should take into account the following general rules: 1. operations should be enclosed as much as possible to prevent contaminating large volumes of air 2. high velocities and cross-drafts which may significantly increase contamination and dust loading should be avoided 3. the rate of air volume withdrawal from the hood should be greater than the rate of generation of contaminated gases in the enclosure 4. wet chemical operations, e.g., wet digestion or solvent treatment, should be separated from dry operations. If possible, separate enclosures should be designated for these activities 5. radioactive aerosols should be removed from the air stream as close to the enclosure as possible. This prevents contamination of other equipment and the exhaust ductwork 6. the value or the need for accountability of material in use in the enclosure may require that the design allow even the smallest chips or turnings to be collected 7. depending on the combustibility of the material in use, a fire protection system or an inert atmosphere may be required 8. accessibility and ease of decontamination of the enclosure and the ductwork must be considered in the design 9. the exhaust fan must be located to ensure that ductwork within the building is maintained a t a negative pressure 10. if compressed gas piping systems or cylinders are to be introduced into the hood, appropriate ventilation systems should be designed to overcome any potential pressurization that might occur from these systems

For laboratories or work areas in which high activity alpha- or beta-emitting radionuclides are being handled, the use of gloved boxes should be considered. Air locks should be incorporated into the design; they should be exhausted if they open directly into the room.

4.6 RADIOACTIVE MATERIAL WASTE MANAGEMENT /

35

Ventilation systems associated with hoods or gloved boxes containing radioactive materials are often operated continuously to minimize the potential for release of these materials. Where necessary, the design should incorporate a standby fan and backup electrical power to ensure continued operation of the system despite component or power supply failures. Appropriate indications should be provided to warn of such failures.

4.6 Radioactive Material Waste Management

To ensure ease of cleanup, the surfaces of floors, walls, fixtures, equipment and work surfaces in areas where radioactive materials may be found should be protected against penetration by contamination. In the selection of materials, consideration must be given to the chemical and physical environment in which the surface materials will function. If porous materials, like concrete, are used in construction, exposed surfaces should be faced with nonporous materials or painted with several layers of nonporous coating (ASTM, 1991).Outer layers of strippable paint may be appropriate. Cracks or holes in surfaces and sharp corners should be eliminated. For some applications, it may be better to design for the use of disposable materials and equipment rather than rely on fixed installations that must be cleaned repeatedly. When deciding which strategy to adopt, the issue of waste generation and management must also be considered. Sinks and drains for radioactive liquid waste should be provided for cleanup in radioactive work areas. Holding and sampling tanks, as well as processing or radioactivity removal systems, may be required for contaminated waste drains and sinks to ensure that radioactive effluents do not exceed permissible levels. Piping systems should be designed to minimize connections between clean and potentially contaminated systems. Discussion of the design of radioactive waste systems for major facilities like power reactors or processing plants is beyond the scope of this Report. To reduce unnecessary exposure, radioactive materials should be stored in areas separate from work places. Ventilation should be provided for storage areas for radioactive material when airborne releases are possible, and access to these areas should be controlled. See Section 8.5 for a discussion of the storage of solid radioactive waste. Other aspects of radioactive waste management are addressed in Section 8.

36 I

4. FACILITY DESIGN

4.7 Instrumentation and

Access Control Systems The electrical design for the facility should include wiring, outlets and other components that are adequate to support initial and projected loads. Where necessary for safety and maintenance of access control, record keeping, and needed computer functions, uninterruptible or backup power supplies should be provided. The design should include adequate fixed monitoring instruments (see Section 4.3) and outlets for portable equipment and instrumentation, such as counting equipment for radioactive samples. Interlocks and the associated circuits that are part of access control and warning systems should be designed to be "fail safe." The controls, relays and wiring should be installed in secure and protected conduits to prevent tampering and unauthorized modification. Computer-controlled programmed logic systems must be protected against accidental and unauthorized changes. Additional design information and recommendations are contained in NCRP Report No. 88 (NCRP, 1986).

4.8 Nuclear Criticality Safety When facilities may contain amounts of fissile material (e.g., materials that could pose a potential criticality hazard, the prevention of inadvertent criticality becomes an important design consideration. Methods of ensuring criticality safety include control of material quantity, form, geometry and the presence of potential neutron moderating, reflecting and absorbing materials. Further discussion of criticality safety design can be found in the following references; Knief (19931, 0fDell(1974),Paxton (1989), and Paxton et al. (1964). 2 3 3 ~ 2, 3 5 ~ 2, 3 9 ~ uor) quantities of other

4.9 Recommended Additional Reading Guide for Nuclear Criticality Safety in the Storage of Fissile Materials, American National Standards InstituteIAmerican Nuclear Society 8.7-1975 (R1987), American Nuclear Society, La Grange Park, Illinois, 1975. Site Selection and Design of an Independent Nuclear Fuel Facilities - Spent Fuel Storage Installation, American National Standards InstitutelAmerican Nuclear Society 2.19-1981 (R1990), American Nuclear Society, La Grange Park, Illinois, 1981.

4.9 RECOMMENDED ADDITIONAL READING

/ 37

Nuclear Criticality Safety in the Storage of Fissile Materials, U.S. Nuclear Regulatory Commission Regulatory Guide 3.43, U.S. Government Printing Office,Washington, 1979.

5. Orientation and Training 5.1 General Principles In keeping with the specific requirements of regulations, licenses, permits and the objectives of good radiation safety practice, employers are responsible for conveying safety policies and procedures of their organizations to visitors and workers to safeguard their health and well-being and to ensure that their activities have minimum negative impact on the environment and public health. To meet these responsibilities, organizations should establish radiation safety orientation and training programs that include opportunities for all workers to receive repeat training a t appropriate intervals. Radiation safety policies and procedures should be integrated into the overall safety program of the organization. The depth and breadth of training needed varies with the job requirements and responsibilities of each individual. Factors that influence the depth of training include the potential for radiation exposure, complexity of tasks to be performed, degree of supervision provided in performing the tasks, amount of previous training, and degree to which the trainees will instruct or supervise others. Workers who need specialized radiation safety skills require extensive and ongoing in-depth training. Ofice workers in a facility where radioactive material or radiation sources are used may require only periodic informal training. The duties of the individual will dictate the extent of training required. Training programs should include opportunities for questions and discussions, and participants should be encouraged to contribute to such discussions. Printed handout materials should be provided for attendees to study after finishing the training session. Testing should be incorporated into the training program as needed to assure adequate understanding of the material presented. Education and training should be developed by individuals with educational backgrounds and experiences appropriate to the safety policies and procedures to be discussed. Trainers must be able to respond to questions and problems that might arise from everyday

5.2 DESIGN OF A GENERAL TRAINING PROGRAM

/ 39

experiences in the workplace and should have s d c i e n t experience in teaching methods to make the training session beneficial to all participants. Records of training programs presented, course curricula and attendance records should be maintained by management. Records of training participation should be entered into the employee files and linked to the course record files (NCRP, 1992).

5.2 Design of a G e n e r a l T r a i n i n g P r o g r a m The process of designing and developing a radiation safety training program is described in NCRP Report No. 71 (NCRP, 1983a). The fundamental steps in this process are: 1. a job evaluation is conducted to determine the knowledge, skills and attitudes necessary to perform a task at the desired level of competence 2. training analysis is performed, including development of training objectives, establishment of testing criteria, and determination of the course structure 3. training materials are developed 4. an evaluation plan is designed along with the training materials 5. training is provided 6. evaluation and feedback are obtained to ensure that the original standards of performance are met and maintained Education in the fundamental knowledge and skills required for individuals who may be occupationally exposed to radiation or radioactive material can be provided in a number of ways. To educate a large number of people, group instruction using lectures and audiovisual material is satisfactory. Individual study can be personalized to the needs of the trainee and can be decentralized, unscheduled, self-paced, self-evaluated and self-reinforced. Many educational media can be employed in a self-instruction program, including traditional textbooks, manuals of programmed instruction, slide-tape series, videotapes, and interactive computer-based learning programs. Limitations of individual study include limited opportunities for interaction with the instructor or other students, and the need for considerable self-motivation. Radiation safety education should include, as appropriate, treatment of the following topics:

40 / 5. ORIENTATION AND TRAINING 1. 2. 3. 4. 5. 6. 7.

8.

9. 10. 11. 12. 13.

radioactivity and radioactive decay characteristics of ionizing radiation radioactive materials and radiation sources relative importance of the various sources of exposure radiation measurement effects of exposure to radiation risks associated with occupational and nonoccupational exposures effective dose limits and the application of the principle of ALARA special considerations in the exposure of women of reproductive age [see Section 10 of NCRP Report No. 116 (NCRP, 1993b)l mode of exposure-internal, external effective dose determinations warning signs and alarms basic protective measures-time, distance, shielding

In addition to the above, training in the specialized skills that are required for working with radioactive material or radiationproducing devices should be provided in a radiation safety training program that is specific for the radiation hazards a t the facility and includes the following topics as appropriate: 1. basic radiation survey instrument use and limitations 2. radiation monitoring programs and procedures 3. contamination control, protective clothing and equipment, workplace design 4. personal decontamination 5. emergency procedures 6. warning signs, alarms, posting requirements 7. responsibilities of employees and of the organization 8. interaction with radiation safety staff 9. specific procedures for maintaining radiation exposures ALARA 10. regulations, license and permit requirements 11. radioactive waste policies

Continuing education and training must be provided for the staff that are responsible for the radiation safety program and that handle radioactive material or operate radiation-producing devices. Management must support and encourage the attendance of radiation safety staff a t continuing education programs and

5.3 SPECIFIC TRAINING REQUIREMENTS

/ 41

scientific meetings to remain current in the field. Refresher courses in the topics listed above should be provided for workers on a regular schedule. 5.3 Specific Training Requirements

Occupationally exposed individuals may need additional information in the form of specific training on the proper procedures for their individual jobs. The responsible supervisor, with assistance from the RSO, must assure that this training is provided. The information provided should emphasize the potential radiation hazards, radiation control devices, and procedures unique to the individual's work area, including topics such as monitoring techniques and waste disposal procedures and worker concerns. Particular attention should be given to the need and procedure for preplanning all jobs that might result in release of radioactive material or exposure to radiation. Procedures, permits and instructions should be prepared and reviewed a t pre-job briefings and meetings. Special training in contamination control, dose control, and dose reduction may be appropriate for individuals who are responsible for design or modification of facilities. When equipment or techniques involve a risk of significant radiation exposure or release of radioactive material, training sessions should be supplemented with practical training mock-ups designed to allow the employee to develop and demonstrate proficiency in use of the equipment or techniques. Examples include x-ray machine operation, pipetting techniques, gloved box operation, experimental procedures, use of protective clothing, use of respiratory protective equipment, and use of portable radiation monitoring instruments. Proficiency demonstrations should be part of a formal qualification program before a new employee is allowed to begin work and repeated a t appropriate intervals. This training should be documented. Training on standard procedures for recognizing and responding to radiation emergencies (NCRP, 1980; 1991a) is an essential part of the program. All credible emergency situations should be reviewed, ranging from small spills of radioactive materials to major accidents. The employee should understand his or her role in responding to an emergency. Specific actions that might be performed, ranging from immediate evacuation to mitigation of the consequences of an emergency, should be discussed and practiced.

6. External Radiation Exposure Control

This Section provides guidance for the operational radiation safety program that is required to control exposures to external radiation sources. Guidance is included for dose control techniques, workplace monitoring, personal dosimetry, and protective clothing. The use of engineering controls, work planning, and monitoring of the workplace are emphasized.

An external radiation exposure control program shall be established when there is a possibility for workers to be occupationally exposed or for members of the public to receive exposure from facility operations. The ICRP has used the expression "occupationa1 exposure" to characterize exposures incurred a t work as the result of situations which can reasonably be regarded as being the responsibility of management [see paragraph 134, ICRP Publication 60 (ICRP, 1991a:Il. The formality of the program is clearly a function of the dose level. At low doses, below about 1mSv y-l for workers, the program may be limited to broad dose assessments. At higher dose levels, the level of training, program oversight, etc. should be increased. For members of the public, expected exposures in excess of 0.01 mSv y-l would, in a similar way, suggest that formality be increased with the level of projected dose. Results of the monitoring of external exposures can establish the general conditions of the workplace, whether these conditions are under satisfactory control, and whether operational changes have affected the work environment. The frequency and extent of the measurements should be reviewed periodically and adjusted to meet the needs of the program. Other aspects of the program such as facility design and operating procedures should also be reviewed periodically.

6.2 RADIATION DOSE CONTROL TECHNIQUES

1 43

6.1 Radiation Dose Controls 6.1.1

Limits

The objective of a n external radiation exposure control program is to maintain occupational radiation doses, and radiation dose to the public, in accordance with authoritative exposure standards. Recommended occupational radiation dose limits have been promulgated by the NCRP (199313)and the ICRP (1991a). These radiation dose limits include both internal and external exposures and are specified as annual limits of either effective or equivalent dose as appropriate and in the case of the NCRP, a lifetime limit is specified. Current recommendations of the NCRP are listed in Table 1.1. In addition, regulatory bodies have established limits pursuant to their legal authority. 6.1.2

Administrative Dose Guidelines

Administrative dose guidelines (reference levels) should be established to reduce the potential for individuals to exceed the recommended dose limits. An effective external radiation exposure control program will ensure that doses to occupationally exposed individuals are maintained within administrative dose guidelines and that individual doses are maintained ALARA (see Section 2) for the work to be performed. Administrative dose guidelines should not be exceeded without proper review and approval except during emergencies. These guidelines may be established for a particular task, a portion of the year, or an entire year (see Section 6.2.3). 6.2 Radiation Dose Control Techniques

In facilities (e.g., research laboratories, radiopharmacies, radiography centers) where radioactive materials are handled or where radiation-producing equipment is used, the building, equipment, safety procedures, and work-planning activities should be designed and implemented to ensure that radiation doses are maintained within recommended limits and are ALARA. Engineered controls should be the primary means for controlling external radiation doses. These include distance and shielding, remote handling equipment and interlocks. Admmistrative controls such a s safety procedures, radiation work permits (RWP), and radiation

44 / 6. EXTERNAL RADIATION EXPOSURE CONTROL

monitoring and surveys should be a secondary means for controlling external doses, but are a necessary part of the program. The components of an external radiation dose control program are discussed in the remainder of this Section.

6.2.1

Time, Distance and Shielding

The principal controls used to reduce occupational external doses from radioactive materials or radiation-producing equipment to acceptable levels are separating the radiation source from occupied areas, utilizing radiation shielding, and minimizing the time spent in elevated radiation fields. Source storage and waste storage areas can be located remotely from occupied areas to minimize the shielding required. Design considerations for radiation shielding (see also Section 4.4 for additional detail) include the type, energy, intensity and orientation of the primary radiation source, the materials used for shielding, the use factor of the source, the occupancy time of individuals in the area outside the shield, and the potential changes in these parameters. The factors that must be considered in selecting material for radiation shielding include the type and intensity of the radiation source, the need for structural integrity, space limitations, the need to be able to move the shield for operational reasons, and the cost. Lead, steel, concrete and water are effective shields for x and gamma rays. Beta radiation sources can be effectively shielded by surrounding the source with low atomic number material that completely absorbs the beta radiation without producing excessive bremsstrahlung. Depending on the neutron energy spectrum, materials that have a high hydrogen content, such as concrete, water and polyethylene, can be effective shields for neutrons, and are used alone or in combination with high density materials such as iron or lead. However, care must be taken to shleld adequately the gamma rays that are associated with neutron capture. It is necessary to design radiation shielding such that doses to exposed individuals are not expected to exceed the annual dose limit, t a h n g into account design, construction and measurement uncertainties. The design criterion for dose depends on whether the facility is accessible to the general public. In most cases it is possible to achieve this design level without incurring unreasonable shielding costs (NCRP, 1976a). Shielding design should also incorporate the application of the ALARA principle. Specific guidance for selecting materials for neutron shielding is contained in NCRP

6.2 RADIATION DOSE CONTROL TECHNIQUES

/ 45

Report No. 38 (NCRP, 1971) (recommended additional reading is given in Section 4.4). 6.2.2

Access Control and Alarm Systems

Access control systems and alarm systems are designed to w a n workers and visitors in a facility of the potential for exposure to radiation and to reduce the likelihood of that exposure. Access control systems are designed to prevent inadvertent or unauthorized access to areas where elevated dose rates can exist. Alarm systems are designed to alert people if the radiation dose rate increases above a predetermined level. An alarm system may also provide instructions and it may initiate mitigating actions. A detailed discussion of access control and alarm systems, and the criteria for their selection and use is presented in NCRP Report No. 88 (NCRP, 1986). A brief summary is provided here relevant to external dose control. The degree of sophistication needed for access control and alarm systems is determined by the potential for inadvertent exposure. For example, the access control system for a calibration facility in which a small source is exposed and the radiation hazard is relatively minor may be a simple warning sign and a rope barrier. However, the access control system for a n accelerator, where life-threatening levels of radiation may be present, would be more elaborate and include interlocked barriers, audible and visual warnings, and explanatory signs. The access control system can include signs, visual and audible signals, physical barriers, interlocks, run-safe switches, emergency shutdown switches, pre-startup notification and search procedures, administrative procedures, and special instructions. Signs that are used as part of a n access control system should have a standard format that includes: 1. a heading that indicates the degree of hazard, e.g., Caution or Danger 2. a statement of the type of hazard 3. the standard radiation symbol (ANSI, 1989a) 4. a brief statement of instructions

CAUTION should be used on signs where the risk is limited to receiving unnecessary radiation doses or receiving radiation doses that could exceed permissible limits. DANGER should be used for areas where the risk is of immediate injury or death.

46 / 6. EXTERNAL RADIATION EXPOSURE CONTROL

Visual signals should be flashing or rotating beacons that are magenta, safety purple, red, or for dimly lit areas, black on white. A sign a t each signal should indicate the purpose of the light and what actions are required. Audible signals should be loud enough and distinctive enough to be heard over the ambient noise level, but not so loud or irritating that workers are tempted to disable them. A pulsed chime is often found to be satisfactory. Barriers can range from something as simple as a rope to something as complex as an interlocked concrete door. However, the barrier should not prevent rapid egress in an emergency. The intrusion protection provided by a barrier should be commensurate with the potential radiation exposure hazard. Often shielding around a radiation source may be an effective access barrier. An interlock is a device on an access control barrier that automatically prevents radiation or reduces the dose rate when the barrier is opened and access to the area is permitted. Interlocks should be used as part of an access control system for areas in which there is the potential for an individual to receive an effective dose exceeding 50 mSv (Table 6.1). Interlocks should be "fail-safe";that is, in their most likely failure modes, they will prevent high radiation levels in potentially occupied areas. Run-safe switches and search procedures should be used in large or complex radiation exclusion areas to prevent the presence of radiation until the areas are known to be unoccupied and secured. Emergency shutdown switches should be installed in radiation exclusion areas to prevent or terminate prompt radiation if the area is inadvertently occupied. Pre-startup notification is a visual and audible indication that radiation is imminent in an exclusion area. The notification should consist of a dimming of the lights in the area to alert any persons who might remain in the area following the search. This should be followed by an oral announcement or a unique sound, or both, to indicate that radiation is imminent in the area. Alarm systems are normally activated directly by radiation sensors that trigger a warning device when a preset radiation level has been reached. Such systems include area monitors or criticality accident alarm systems. Area monitors are used to monitor ambient radiation levels in potentially occupied areas and will activate an alarm when a predetermined level of radiation has been reached or exceeded. They are generally used for penetrating radiations such as x and gamma

6.2 RADIATION DOSE CONTROL TECHNIQUES

/ 47

TABLE6.1-Alarm and access control systems as a function

of effectivedose that an individual might receive from an inadvertent exposure.a Effective Dose Category (mSv)

Alarm System

Access Control System

51

None

Signs and ropes

>1 to 50

Visual and audible

Signs and barriersb

>50 to 250

Visual and audible

Interlocks, barriersb and signs

>250 to 1,000

Visual and audible

Interlocks, b a ~ r i e r ssigns, ,~ lights and audible alarms

>1,000

Visual and audible

Interlocks, barrier^,^ signs, lights, audible alarms, run-safe and emergency off switches

From NCRP Report No. 88 (NCRP, 1986). Barriers in these dose categories normally include shielding materials such as concrete, steel, earth and lead to reduce the radiation levels in occupied areas to acceptable levels. a

rays or neutrons; however, with suitable detector design they can be used for detecting beta-particle radiation. Criticality accident alarm systems are a special class of area monitors that detect the ionizing radiation from an accidental nuclear criticality. Such devices must respond to bursts of high intensity neutron or gamma radiation associated with a criticality accident. The selection of access control and alarm systems is based on the potential effective dose that an individual might receive in the event of an inadvertent exposure. In selecting these systems, the potential equivalent dose rate, the portion of the body exposed, and the potential exposure time should be used to determine a potential effective dose. General guidance for this selection process is shown in Table 6.1, which is extracted from NCRP Report No. 88 (NCRP, 1986).

48 /

6.2.3

6. EXTERNAL RADIATION EXPOSURE CONTROL

Radiation Safety Procedures and Radiation Work Permits

Work that involves the use of radiation-producing equipment or significant quantities of radioactive materials during which there is a potential for exceeding the administrative dose control limits should be authorized by a safety procedure or RWP. Safety procedures describe and authorize tasks that are repetitive and will be performed over an extended time period. RWPs describe and authorize specific tasks that require special radiation safety controls. Safety procedures should be reviewed and updated periodically. RWPs should be terminated when the task has been completed. Both safety procedures and RWPs should contain information for controlling both external and internal exposures to ionizing radiation and contain the following information (see Section 7.2.2 as appropriate):

1. a description of the work that is authorized by the safety procedure or RWP and identification of the work procedures to be used 2. a description of the potential hazards that will be encountered in performing the work, including potential radiation dose rates, identification of the sources of radioactive material, the potential radioactive contamination levels, and the potential for intake of radioactive material 3. the individuals responsible for making sure that the work activities are conducted in accordance with the safety procedure or RWP 4. the safety controls and procedural safeguards that are necessary to prevent or limit the exposure including requirements for protective clothing respiratory protection external dosimetry measurements to detect internally deposited radionuclides radiation surveys worker time and dose limitations limiting conditions on the radiation or contamination levels above which the safety procedures or RWP is void health physics or radiation safety coverage that is required during the task

6.3 EXTERNAL RADIATION DOSIMETRY

/ 49

5. worker qualifications including any specialized training that is required 6. actions to be followed in the event of a n emergency RWPs should also include a requirement that each person who enters the work area acknowledge by signature or by electronic means that he or she has read and understands the permit. 6.2.4

Exposure Planning and Dose Reduction Activities

Engineered controls can be used effectively to reduce the immediate hazard to the workers. However, these controls add complexity to the operation and require additional monitoring to ensure that they function as designed. One of the components of a work planning activity is to evaluate the potential external radiation doses to workers and to consider engineering and administrative controls for each task that will maintain the individual and collective doses ALARA (see Section 2). The extent and formality of activities for exposure planning and dose reduction should be commensurate with the hazard of the task and the potential dose savings. Careful planning of work activities coupled with a pre-work briefing, training and the selection of appropriate equipment and protective devices can do much to control the exposure of workers to radiation and other hazards. Specific items can be documented on the RWP to ensure that the information transmitted to the workers is unambiguous. Post-job reviews can provide valuable information that can be used to reduce future exposures from similar tasks. 6.3 External Radiation Dosimetry 6.3.1

Personal Monitoring

The assessment of the radiation dose received from external sources by occupationally exposed individuals serves two purposes. First, it provides the information necessary to compare a worker's occupational dose with both the dose limits and administrative dose guidelines. Second, it provides information about the effectiveness of the external dose control program. A detailed discussion of personal monitoring techniques is presented in NCRP Report No. 57 (NCRP, 1978b). A summary is presented here.

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6. EXTERNAL RADIATION EXPOSURE CONTROL

Personal dosimeter devices, including thermoluminescent dosimeters (TLD), film dosimeters, CR-39 polycarbonate foils, pocket ionization chambers, and electronic personal dosimeters, provide an integrated measurement of the radiation dose. Passive dosimeters such as TLD, film and CR-39 require special processing to obtain the dose. Active dosimeters such as pocket ionization chambers and electronic personal dosimeters provide a direct real-time readout. Normally, passive dosimeters are used to provide the primary and permanent record of a worker's radiation dose because they provide an economical and acceptably accurate estimate of the integrated dose from various components of complex radiation fields. Active dosimeters are normally used to provide real-time estimates as the dose accumulates. Electronic personal dosimeters are typically equipped with audible alarms that are activated when a preset dose or dose rate is reached. They are used when a n immediate indication is needed that a dose or dose rate control level is being approached. The radiation dosimetry device that is to be used for the permanent record of worker doses should be capable of measuring the types and energies of the radiation that will be encountered with sufficient accuracy, sensitivity and precision. Its response should be checked periodically and the program used to calibrate and process the dosimeter should be validated by a recognized dosimetry accreditation program. The more complex the radiation field, the more sophisticated are the requirements for the personal dosimeter to measure dose accurately. Provision of personal dosimeters for external exposure measurement should be considered for workers who are likely to receive an annual effective dose in excess of 1mSv. Personal dosimeters should be issued to visitors if they have the potential for receiving a dose that exceeds 25 percent of the applicable dose limit for the public. The length of time that a passive personal dosimeter is worn before it is evaluated should be established on the basis of the potential dose that could be received in the interval, the capability of the dosimeter to integrate accurately the radiation dose information over extended periods of time, and the need to obtain timely information about radiation exposures. Monthly or quarterly personal dosimeter exchange cycles are common for workers. If the radiation fields are reasonably uniform, a single personal dosimeter worn a t the appropriate location on the trunk of the body is usually sufficient to properly characterize the worker's radiation dose. However, if the radiation fields are not uniform or if the head

6.4 MONITORING AND SURVEILLANCE PROGRAM /

51

or extremities are likely to be exposed to higher radiation fields than the trunk, additional dosimeters may be required (NCRP, 1995a). Nuclear criticality accident dosimeters should be issued to workers when fissile material ( 2 3 3 ~2 ,3 5 ~239Pu) , is handled in sufficient quantities that an uncontrolled chain reaction could occur. Because the neutron and gamma radiation doses from criticality accidents can range from grays to hundreds of grays, specialized dosimeters are required to measure these radiation doses accurately.

6.3.2

Dose Assessment

Passive personal dosimeters typically contain a variety of filters that allow the readings on the dosimeters to be interpreted to estimate the dose from the various components of the radiation field. Care must be taken to calibrate the dosimeter to respond with acceptable accuracy to the radiation fields that will be encountered in the workplace. The dosimeter selected should be able to measure the personal dose equivalent (ICRU, 1993),H (dl a t a depth, d, of 10 rnrn for penetrating radiation and also at t i e depth of 0.07 mm if the monitored individual is expected to be exposed to nonpenetrating radiation. Hp(lO)and Hp(0.07) are also referred to as the "deep dose" and "shallow dose," respectively (ANSI, 1993). If a worker is exposed to nonuniform radiation fields, the determination of the actual dose received can be a very complex process. In general, for doses well below the recommended limits, a single measurement taken at the appropriate location on the surface of the trunk, provides enough information to estimate the dose with sufficient accuracy. However, as the dose approaches, or exceeds, the annual limit, it may be necessary to determine the effective dose more precisely. This can be done using the recommendations and information provided in NCRP Report No. 122 (NCRP, 1995a).

6.4 Monitoring and Surveillance Program 6.4.1

Radiation Surveys

Radiation surveys should be conducted in areas where the potential exists for exposure to external radiation fields in order to:

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6. EXTERNAL RADIATION EXPOSURE CONTROL

1. characterize the radiation field so that it can be properly posted and controlled 2. provide the information required for planning work activities to maintain the external radiation exposures a t levels

ALARA 3. ensure the prompt discovery of changed radiation fields caused by changing conditions Radiation surveys should be conducted a t a frequency that is commensurate with the potential for changes in the radiation fields and the potential magnitude of the changes. Routine radiation surveys are conducted a t fixed intervals and a t fixed locations to document the field and to determine whether there have been any unexpected changes in the external radiation field levels. Nonroutine surveys should be performed to evaluate radiation fields that have not been previously measured and when there is an expected change in the radiation field, for example: 1. during the initial operations of newly installed radiation-producing equipment or radiation sources 2. following the modification of radiation-producing equipment or radiation sources 3. following any modification of the shielding around a source of external radiation 4. following a n incident in which an elevated external radiation exposure is suspected or has occurred

The instrumentation that is used to perform radiation surveys should be capable of measuring accurately the types of radiation, a t the dose rates and under the environmental conditions that may be encountered. The radiation survey results should be documented according to established procedures and should include: 1. a description or drawing showing each measurement location 2. the measured dose rates a t each measurement location 3. the type, the model number, the serial number, and the calibration date of the survey meters used 4. the name and signature of the individual who performed the survey 5. the date and time that the survey was performed

6.5 PROTECTIVE CLOTHING

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6 . comments by the persons making and evaluating the survey that are relevant to the interpretation of the survey data The radiation doses to the workers who perform the radiation survey should be maintained ALARA. 6.4.2

Area Monitoring

Fixed area monitors are typically used to monitor ambient radiation levels in potentially occupied areas. They are particularly useful when the potential exists for a significant increase in the ambient radiation levels. These monitors provide a continuous "radiation survey" a t their predetermined and fixed locations and normally activate a n alarm when a predetermined radiation dose or dose rate is exceeded. They can also be used to supplement or replace personal dosimeters, especially in areas where the annual external effective dose is expected to be less than 1mSv. The design and implementation of an area monitoring program should include, as appropriate: 1. a n evaluation of the radiation to be detected 2. the range of dose rate that may be encountered 3. the location of the detectors relative to the workers and the radiation source 4. the operating and response characteristics of the monitoring system 5. effects of the environment on the monitoring system

6.5 Protective Clothing In most cases, protective clothing is used to avoid getting radioactive contamination on the worker, to prevent the spread of contamination, and, in some instances, to provide protection against external radiation. It is always preferable to shield the radiation a t its source rather than to place a shield on the worker. However, in some instances cost or operational flexibility make it reasonable for the worker to wear clothing that provides radiation shielding. Protective clothing for external radiation is effective primarily against beta particles and x and gamma rays with energies less than 200 keV. The most common types of protective clothing for external radiation are lead impregnated gloves, aprons and vests. The

54 1

6. EXTERNAL RADIATION EXPOSURE CONTROL

density of lead limits its practical thickness in aprons and vests to about 1 mm, which provides attenuation in excess of a factor of three for photons less than 200 keV. The lead thickness in gloves is limited to about 0.03 mm to limit the loss of flexibility. Specific recommendations for medical workers are given in NCRP Report No. 102 (NCRP, 19894. Plastic glasses or face shields can be useful for reducing the dose from beta particles. The thicknesses of the devices should be selected such that they are effective in absorbing the beta particles emitted from the source being handled. A lab coat, or a single or double layer of coveralls that is useful for eliminating the direct contamination of the skin can also provide some protection against beta radiation. The actual protection afforded depends on the thickness of the material and the energy of the beta particles (Farrell and Hudson, 1985; Hudson, 1983; Shonka e t al., 1990). When selecting clothing for protection against external radiation, a n evaluation should be made of the potential increase in working time and radiation dose caused by the loss of dexterity and the added weight of the clothing. It is possible that the anticipated dose reduction provided by the clothing will be more than offset by the concomitant increase in the time spent in the radiation field. Clothing used for protection against external contamination or radiation can become damaged with use. Lead shielding in gloves, vests and aprons can shift with time. Thus, all such protective clothing should be inspected periodically to ensure its continued integrity and usefulness. 6.6 Records

External radiation dose records should be maintained to demonstrate compliance with dose limits and administrative dose guidelines, and to assist in the evaluation of the effectiveness of the external dose control program. Recommendations for the content of occupational dose records are given in NCRP Report No. 114 (NCRP, 1992) and include the following: 1. the annual occupational radiation dose for each year that the worker was monitored, including the doses received during previous employment 2. the cumulative occupational radiation dose received by the worker

6.7 RECOMMENDED ADDITIONAL READING / 55

3. the respective doses for those parts of the body that were monitored separately 4. any significant accidental exposures including the circumstances of the accident, the cause of the accident and the evaluation of personal doses 5. the circumstances of any planned special exposures, and in particular, those that exceeded the dose limits for normal operations 6. any special evaluations of personal dose that resulted in a n adjustment to the measured dose to obtain an assigned dose 7. a description of the dosimetry system used and a discussion of the calibration methodology; 8. evaluations for the application of the ALARA principle

In addition to the records maintained for the personal dosimetry program, records should be maintained of the external radiation surveys that are performed. The information that should be included is listed in Section 6.4.1. 6.7 Recommended Additional Reading

"Control system dependability considerations encompass availability, integrity, and security," Lynott, F., Control Eng. 29, 22-24, 1982. Dose Control at Nuclear Power Plants, NCRP Report No. 120, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1994. Radiation Protection for Medical and Allied Health Personnel, NCRP Report No. 105, National Council on Radiation Protection and Measurements, Bethesda, Maryland, 1989. "Redundancy used in fail safety alarm design," Ida, E.S., Control Eng. 30, 89-91, 1983. Technological Conszderations in Emergency Znstrumentation Preparedness, Phase ZZ-D-Evaluation, Testing, and Calibration Methodology for Emergency Radiological Znstrumentation, Bramson, P.E., Andersen, B.V., Fleming, D.M., Kathren, R.I., Mulhern, O.R., Newton, C.E., Oscarson, E.E. and Shelby, J.M., Report BNWL-1991, Battelle Pacific Northwest Laboratory, Richland, Washington, 1976. Tutorial Overview of Design, Fabrication and Testing of Radiation Hardened Custom ZCs, CMOS, Gwyn, C.W., Tutorial Short Course, American Nuclear Society, Winter Meeting, Washington.1982.

7. Internal Radiation Exposure Control

This Section provides guidance for the operational radiation safety program that is required to control exposures to internally deposited radionuclides. Guidance is included for contamination control, air monitoring, personal monitoring, bioassay, and respiratory protection. The use of engineering controls, personal protective equipment, monitoring of the workplace, and an effective bioassay program are emphasized. Although dose assessments in a general sense are required for any occupational exposure to airborne radionuclides, a formal internal radiation exposure control program should be established when there is a possibility for workers to encounter radioactive material in quantities that might lead to intakes during a year that would result in a committed effective dose in excess of 1mSv. The purposes of this program are to control the exposure ofworkers and the public under normal and abnormal situations, to provide a check on the effectiveness of engineering controls, to comply with administrative and regulatory limits, to ensure that exposures are ALARA, and to determine the quantity of material that is deposited in the body and the resulting doses should an intake occur. Results of the monitoring that is a part of this program can establish the general conditions of the workplace, whether these conditions are under satisfactory control, and if operational changes have affected the work environment. The frequency and extent of the measurements should be reviewed periodically and adjusted to meet the needs of the program. Other aspects of the program such as facility design and operating procedures should also be reviewed periodically and adjusted to meet the changing needs of the program.

7.2 CONTAMINATION CONTROL PROGRAMS /

57

7.1 Radiation Dose Controls 7.1.1

Limits

The objective of an internal radiation exposure control program is to maintain occupational radiation doses in accordance with authoritative exposure standards and ALARA. Recommended occupational radiation dose limits have been promulgated by the NCRP (1993b) and the ICRP (1991a). These radiation dose limits apply to the sum of the effective dose received from external exposures and the committed effective dose received from exposures to internally deposited radionuclides. Current recommendations of the NCRP are listed in Table 1.1.I n addition, regulatory bodies have established limits pursuant to their legal authority. 7.1.2

Administrative Exposure Guidelines and Reference Levels

Administrative exposure guidelines should be established to reduce the potential for workers to exceed the recommended exposure guidelines and to ensure that individual and collective doses are maintained ALARA (see Section 2) for the work to be performed. An internal radiation control program should ensure that doses to occupationally exposed workers are maintained within administrative dose guidelines. Administrative exposure guidelines should not be exceeded without proper review and approval except during emergencies. These guidelines may be established for a particular task, a portion of the year, or an entire year. The annual reference level of intake and the derived reference air concentration have been defined by the NCRP (1993b). The proper use of these reference levels in monitoring for compliance with the NCRP dose limits is discussed in NCRP Report No. 116 (NCRP, 1993b). The NCRP has also recommended the use of reference levels for design and control purposes (NCRP, 1993b). These reference levels can be calculated using standard metabolic models and computer codes designed for that purpose (ICRP, 1988; Skrable et al., 1987). 7.2 Contamination Control Programs

Although usually not a significant risk to workers, contamination of facilities, equipment or people occurs in many operations involving radioactive material. Contamination control of routine

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7. INTERNAL RADIATION EXPOSURE CONTROL

operations is normally accomplished through containment of the radioactive material in chemical hoods, gloved boxes, hot cells, or the use of area exclusion, protective clothing, etc. In addition, procedures and work planning play a significant role in reducing the possibility of a contamination incident. However, there are a nwnber of routine operations which, through maintenance activities, system leaks, or accidents have the potential to contaminate facilities, equipment or people. These operations range from simple (e.g., chemical manipulations involving a radioactive solution,) to complex (e.g., removal of radioactive materials h m a hot cell). In most cases, contamination should be controlled, and removed as soon as possible. The contaminated area or equipment should be marked and posted immediately. Nonessential persons should be moved out of the area until decontamination has been completed. Usually simple cleaning techniques and procedures are adequate for most decontamination tasks. Spills and contaminated areas should be cleaned from the outer region inward to reduce the possibility of further spread of the contamination. After cleaning, the area or equipment should be surveyed to ensure that all the contamination has been removed. In some cases, special detergents or equipment are required, e.g., sand-blasting, hydro-lasing, and other techniques are used to remove contaminated materials from some surfaces. If materials, equipment or areas cannot be decontaminated, they should be clearly marked to indicate the extent of contamination, the radionuclides involved, the estimated activities, and any special precautions to be taken. Items such as contaminated tools should be clearly marked and procedures should be established for the control and use of these items. Efforts should be made to control and, where possible, reduce the extent of contamination in facilities. Reducing or eliminating contaminated areas results in cost-saving by reducing the need to wear protective clothing and other devices in the area. In addition, this effort increases the efficiency of workers, requires less radiation protection oversight, and reduces lost time devoted to donning and removing anti-contamination clothing and gear. In facilities where radioactive materials are handled, the building, equipment, safety procedures, and work-planning activities should be designed and implemented to ensure that the potential for dispersal of radioactive material is minimized. The primary means for controlling contamination should be engineered controls such as containment, ventilation systems, air filtration, and access control. Administrative controls such as safety procedures, RWPs,

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and monitoring and surveys should be a secondary means for controlling contamination of the workplace or the environment. The components of a contamination control program are discussed in the remainder of this Section. Handling of contaminated persons during a n accident or an emergency is outside the scope of this discussion. There are a number of documents which address this subject (IAEA, 1988; ICRP, 1978; NCRP, 1980).

7.2.1 Access Control and Alarm Systems Access control systems and alarm systems are designed to warn workers and visitors in a facility of the potential for exposure to radiation or radioactive materials and to reduce the likelihood of that exposure. Access control systems are designed to prevent inadvertent or unauthorized access to areas in which airborne or surface contamination may result in an intake of radioactive material. Alarm systems are designed to alert people to the presence of airborne contamination. An alarm system may also provide instructions and it may initiate mitigating actions. A detailed discussion of access control and alarm systems, and the criteria for their selection and use is presented in NCRP Report No. 88 (NCRP, 1986). A brief summary relevant to internal exposure control is provided below. The degree of sophistication needed for access control and alarm systems is determined by the significance of potential airborne radioactive materials and the potential for inadvertent exposure. For some areas a simple warning sign and a rope barrier may be sufficient. For other areas the access control system may be more elaborate and include interlocked barriers, audible and visual warnings, and explanatory signs. The access control system can include signs, visual and audible signals, physical barriers, emergency shutdown switches, administrative procedures, and special instructions. Signs that are used as part of an access control system should have a standard format that includes: 1. a heading that indicates the degree of hazard, e.g., Caution or Danger 2. a statement of the type of hazard 3. the standard radiation symbol (ANSI, 1989a) 4. a brief statement of instructions

CAUTION should be used on signs where the risk is limited to receiving unnecessary exposure to uncontained radioactive material or exposures that could exceed permissible limits. DANGER should be used for areas where the risk is of immediate injury or death. Visual signals should be flashing or rotating beacons that are magenta, safety purple, red, or for dimly lit areas, black on whlte. A sign a t each signal should indicate the purpose of the light and what actions are required. Audible signals should be loud enough and distinctive enough to be heard over the ambient noise level, but not so loud or irritating that workers are tempted to disable them. Barriers can range from something as simple as a rope to something as complex as a remotely controlled locked door. The intrusion protection provided by a barrier should be commensurate with the potential exposure hazard. Alarm systems are normally activated directly by sensors on continuous air monitors that trigger a warning device when a preset airborne radioactivity level has been reached. The selection of access control and alarm systems is based on the potential exposure that an individual might receive in the event of an inadvertent release of radioactive material. Table 6.1 gives recommendations for alarms and access control based upon the radiation dose that a n individual could potentially receive. Although designed for external radiation dose control, the recommendations can be adapted for controlling exposure that might lead to doses from internally deposited radionuclides. 7.2.2

Radiation Safety Procedures and Radiation Work Permits

Work that involves the generation or use of dispersible radioactive materials should be authorized by a safety procedure or RWP. Safety procedures describe and authorize tasks that are repetitive and will be performed over a n extended time period. RWPs describe and authorize specific tasks that require special radiation safety controls. Safety procedures should be reviewed and updated periodically. RWPs should be terminated when the task has been completed. Both safety procedures and RWPs should contain information for controlling external and internal exposures to ionizing radiation and contain the following information, as appropriate (see Section 6.2.3):

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1. a description of the work that is authorized by the safety procedure or RWP and identification of the work procedures to be used 2. a description of the potential hazards that will be encountered in performing the work, including potential radiation dose rates, identification of the sources of radioactive material, the potential radioactive contamination levels and the potential for intake of radioactive material 3. the individuals responsible for making sure that the work activities are conducted in accordance with the safety procedure or RWP 4. the safety controls and procedural safeguards that are necessary to prevent or limit the exposure including requirements for protective clothing respiratory protection internal and external dosimetry radiation surveys worker time and dose limitations limiting conditions on the radiation or contamination levels above which the safety procedures or RWP is void health physics or radiation safety coverage that is required during the task 5. worker qualifications including any specialized training that is required 6. actions to be followed in the event of an emergency

RWPs should also include a requirement that each person who enters the work area acknowledge by signature or by electronic means that he or she has read and understands the permit.

7.2.3

Exposure Planning and Dose Reduction Activities

Engineered controls can be used effectively to reduce the immediate hazard to workers. However, these controls add complexity to the operation and require additional monitoring to ensure that they function as designed. An evaluation of the potential intake of radioactive material should be made to determine the extent of procedural and engineered controls that are required to maintain exposur6s and doses ALARA. The extent and formality of activities for exposure planning and dose reduction should be commensurate with the potential dose savings.

62 / 7.INTERNAL RADIATION EXPOSURE CONTROL

Careful planning of work activities coupled with a pre-work briefing, training and the selection of appropriate equipment and protective devices can do much to control the exposure of workers to radiation and other hazards. Specific items can be documented on the RWP to ensure that the information transmitted to the workers is unambiguous. Post-job reviews can provide valuable information that can be used to reduce future exposures from similar tasks.

7.3 Internal Radiation Dosimetry

7.3.1

Personal Monitoring

The assessment of the radiation dose received from internal sources by occupationally exposed individuals serves two purposes. First, it provides the information necessary to compare a worker's occupational dose with both dose limits and administrative dose guidelines. Second, it provides information about the effectiveness of the internal dose control program. Discussions of personal monitoring techniques, general concepts for internal dosimetry and bioassay procedures are presented in various NCRP reports (NCRP, 1978b; 1985a; 1987; 1995a). A summary is presented here. Criteria for selecting persons for participation in a bioassay program should be based on the probability and the severity of the potential exposure. Samples should be obtained more frequently from individuals in the program who are a t a high risk for exposure than from those a t low risk. Professional judgment is required, not only in selecting participants, but also in establishing the most effective bioassay frequency and technique. An effective bioassay program requires that the persons selected to participate and the type and frequency of sampling be carefully planned by someone who has expertise in radiation safety. This individual should be completely familiar with the operations and the workers to be monitored. The general types of bioassay that should be considered in establishing an appropriate sampling process are: 1. baseline or preparatory 2. termination 3. diagnostic 4. routine or periodic

7.3 INTERNAL RADIATION DOSIMETRY

1 63

The baseline bioassay is meant to establish the pre-existing levels of incorporated radionuclides for each worker. The termination bioassay is used to provide a final record of incorporated radionuclides. Diagnostic bioassays are taken to evaluate, in some detail, intake and retention of radioactive material. Information from the diagnostic bioassay is used to determine additional actions that may be appropriate following exposure, such as additional engineered controls or protection equipment. Routine or periodic bioassays are obtained to evaluate the effectiveness of the protection program and to ensure compliance with dose limits and administrative dose guidelines. This sampling must be done on a regular basis and it is important that an appropriate sampling frequency be established. The frequency of routine bioassay should be based on the type of facility, the activities being performed, air sampling results, the potential magnitude of the exposure, and the ability of the bioassay technique to detect an uptake of the radionuclide(s) of concern. Detection should be a t a level that is acceptable for the protection of the health and safety of the worker. This last point requires careful consideration of the radiation dose limits, the use of administrative dose guidelines, the acceptable uncertainty in the estimated intake and the biokinetic model for the radionuclide (particularly the retention). The frequency selected should ensure that a deposition that exceeds the established administrative dose guidelines will not go undetected by the bioassay technique used. Sound judgment by a responsible health physicist is required to establish the appropriate sampling frequency. Additional recommendations for bioassay monitoring programs are found in ICRP Publication 54 (ICRP, 1988). 7.3.2

Bioassay Measurements

There are two general types of bioassay measurements, direct and indirect. The method that is selected depends on the route of entry into the body, the solubility (or transportability) of the material, the metabolism of the material, knowledge of the route of excretion, the sensitivity of the measurement technique, and many other factors. Typically only one bioassay measurement technique is selected for routine monitoring. However, it is common to use as many different methods as are appropriate to evaluate significant exposures. Direct bioassay (often called in vivo bioassay) involves the "direct" measurement of the radioactivity in organs or tissues, or

64 / 7.INTERNAL RADIATION EXPOSURE CONTROL the entire body. This measurement is accomplished by positioning very sensitive radiation detectors near the body and detecting the radiation that escapes the body. This method is used primarily to detect photon-emitting radionuclides. Information on direct bioassay techniques is available (NCRP, 1987; Toohey et al., 1991), as well as information that provides a means for estimating intakes of radionuclides based upon an evaluation of the appropriate bioassay compartment through direct measurements (Lessard et al., 1987). The advantages of direct bioassay include: ; a relatively high accuracy for almost all iphoton emitters ability to measure radioactivity in specific organs (e.g., lungs, thyroid) prompt data acquisition and analysis 4. preclusion of the need for metabolic modkls to estimate the I organ or body radioactivity content 5. simultaneous detection of many radionq!clides 6. preclusion of the need to give and ,handle biological materials

I

The disadvantages of direct bioassay include/ 1. a lower detection sensitivity than excretion analyses (which are not necessarily more accurate) 2. difficulty in discriminating between internal and external contamination 3. a need for elaborate and generally expensive equipment 4. usefulness is limited for radionuclides that emit photons with energy less than 100 keV 5. alpha and beta-particle radiations tire not normally detected 6. calibration sources and phantoms are r

Indirect bioassay (often called in uitro bioassqy) includes a number of techniques that are designed to measure the concentration of radioactive material in biological samples, inclllding urine, feces, exhaled breath, perspiration, saliva, blood and epen hair, fingernail and biopsy samples. A fundamental knowledge:of the metabolism of the radionuclide in the body and the relationbhip of the concentration in the bioassay sample to the quantity inlthe organs and tissues of interest is required to select the apfiropriate bioassay technique. More detailed information on indi~iectbioassay techniques may be found in other sources (Boecker qt al., 1991; NCRP,

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1987).The report by Lessard et al. (1987) is very useful for estimating intakes based on measurements of activity in selected bioassay compartments at known times after the intake. The advantages of indirect bioassay include: 1. large numbers of people can be monitored, usually with minimal interference with the work schedule 2. radiations that are not easily measured by external means (e.g., alpha and beta particles) can be detected 3. external contamination can be excluded 4. in many cases, the cost of the analysis is relatively low 5. detection of the presence of almost all radionuclides is possible with standard analytical techniques and procedures The disadvantages of indirect bioassay are: 1. retention characteristics of the radioactive compound vary among individuals 2. individual variations in uptake and retention are not always known 3. excretion kinetics are not known for all radionuclides 4. some analytical techniques can be time-consuming 5. presence of biohazards and the need to handle biological waste

One approach that can be used to expedite results is to establish a simple, inexpensive procedure to screen the bioassay samples. A more extensive, complex and often expensive procedure can be used to analyze samples that are identified by the screening process as significant or those samples that are obtained as the result of a known exposure. 7.3.3

Dose Assessment

When radioactive materials enter the body, a direct measurement of the radiation dose is not possible. The assessment of effective dose requires an analysis of the type and amount of material taken into the body, its metabolism and excretion from the body. This information is then used to calculate an est.imate of the committed effective dose. To perform this assessment reliably requires that a program be established that specifies sampling frequencies and measurement procedures in accordance with accepted

methodologies and that provides a means to verify the results. The program must also be adequate to acquire the necessary information for the assessment. The information required to assess the internal dose following a n intake of radioactive materials is: 1. the route of entry of the radionuclide or radioactive compound into the body 2. the chemical form of the radioactive compound 3. the metabolism of the elemental form of the radionuclide or the radioactive compound 4. the rate of elimination (biokinetics) of the radioactive compound and its metabolites 5 . the physical properties of the radiations emitted by the radionuclide 6. a n estimate of the body content, organ content, or the magnitude of the intake of the radionuclide

Published calculations of the internal committed dose equivalent to tissues of the body are currently normalized to a unit intake of activity, i.e., Sv B ~ - (EPA, ' 1988; ICRP, 1979; 1991b; 1994). These calculations use averaged metabolic data and are adequate for assessing routine exposures that are well below the effective dose limit. For routine exposures the health physicist must provide a reasonable estimate of the total intake of activity of a particular radionuclide. Estimates of the committed equivalent doses to various tissues and the committed effective dose can easily be made from the published calculations. The assessment of nonroutine exposures to internally deposited radionuclides that approach or exceed the limit on effective dose should be based on the actual metabolism of the material in the exposed individual. 7.4 Monitoring and Surveillance Program

7.4.1

Monitoring for Airborne Radioactivity

There should be a n airborne monitoring program for radioactive materials in those areas where there is a significant potential for airborne contamination. I t is not appropriate to use personal monitoring devices to control internal exposures. Thus, continuously operating samplers equipped with continuous detection devices

7.4 MONITORING

AND SURVEILLANCE PROGRAM / 67

may be needed to detect unexpected airborne contamination and to give a warning of any sudden change in concentration of airborne radioactive material. The location from which the air samples are taken is important for the evaluation of potential exposures to airborne radionuclides. Fixed sampling locations, especially if they are located close to the breathing zones of the workers, may be acceptable for approximating the exposure of workers. However, fixed air samplers can underestimate exposures by factors that range from 100 to 1,000. Therefore, personal air samplers that are worn by the worker are the method of choice for monitoring workers in areas of airborne radioactive material. Surveys for airborne radioactive material should be performed on a regular basis in areas where the potential for airborne radioactive material exists. The frequency of these surveys should be commensurate with the potential for the existence of, or changes in, airborne radioactivity levels. Special surveys for airborne radioactivity should be performed to evaluate conditions that have not been previously measured and when there is an expectation that airborne radioactivity will be present or that airborne radioactivity levels may have changed, for example: 1. any work or operation that involves opening a system that is known to contain (or have contained) radioactive material 2. while conducting cutting, grinding and welding operations on radiologically contaminated materials 3. during initial entry into (and periodically thereafter) any area that is known or suspected to contain airborne radionuclides or significant loose surface contamination 4. immediately following the discovery of a significant spill or spread of radioactive materials 5. whenever respiratory protection devices are used to control the intake of radioactive materials Sources of airborne radioactive material may be identified and evaluated by obtaining grab samples in specific locations. The Samples should be representative of the air breathed by persons in the area. Low volume air sampling should be used to determine radioactivity levels for the protection of workers. High volume air Sampling should be used in situations in which airborne radioactive material concentrations need to be determined rapidly or the specific activity of the material is low.

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7. INTERNAL RADIATION EXPOSURE CONTROL

Continuous air monitors should be used when radioactivity levels can fluctuate and early detection could prevent or minimize exposure. These monitors should be located in areas in which the potential for internal exposure is high and near systems that have a potential for causing rapid increases in airborne radionuclides. Continuous air monitors can be equipped with an alarm system that will indicate an abrupt change or an increase in radioactivity above a preset level. Alarms from these monitors are frequently the first indication that a n intake may have occurred. They may also be the only source of useful data for some radionuclides. A comprehensive air monitoring program is essential to the evaluation and control of the hazard associated with airborne radioactive materials. It provides important information that is necessary to: 1. evaluate the level of airborne contaminants 2. identify and characterize airborne contaminants and, potentially, their source 3. determine the requirements for engineered and administrative controls 4. record long-term trends in the work environment 5. inhcate the continuing effectiveness of existing controls 6 . document worker exposures and long-term collective exposures 7. warn of the deterioration of engineered controls and operating procedures, or of unanticipated releases 8. assist in the evaluation of potential intakes of radioactive material

Assessments of internal dose that are based on air sampling results and the estimate of exposure time are generally less accurate than assessments that are based on bioassay data. Therefore, they should be used for internal dose assessment only when whole body counts and other bioassay information are not available. The air monitoring program should be designed so that the data obtained are related to the operation or process of concern. The program design and implementation should include, as appropriate: 1. an evaluation of the radiations emitted by the radioactive materials to be sampled 2, the range of concentrations that are likely to be encountered

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3. sampler location relative to workers and the source of contamination

4. the portability, operating and response characteristics of the sampling system 5. the chemical reactivity of the contaminant and the temperature sensitivity of the sampling system

For certain applications it may be important to determine the chemical and physical state of the contaminant as well as the aerodynamic characteristics of the particulate material. 7.4.2

Contamination Surveys

Contamination surveys are conducted to establish the extent to which radioactive material is present on surfaces of equipment and in facilities and the extent to which that material may be transferable. Any transfer or resuspension of the material can lead to internal deposition through ingestion or inhalation. The instruments that should be used for such surveys and the associated measurement uncertainties are discussed in Section 10.4. If radioactive contamination is possible on workplace surfaces, simple measures such as restrictions against smoking, eating or drinking in the area will reduce the potential for intakes. In some cases, protective clothing and respirators may be necessary. When contamination is found, decontamination procedures as hscussed in Section 7.2, should be instituted.

7.5 Protective Equipment and Devices 7.5.1

Containment Systems

An effective method to limit the potential for an internal radiation exposure is to provide systems or devices that will keep the radioactive material contained. A containment system can be a simple plastic gloved enclosure, a chemical hood, or a gloved box. Selection of the appropriate containment system should be based on the assessment of the potential for release of material into the working environment and for the intake of the material by workers in the area.

70 / 7. INTERNAL RADIATION EXPOSURE CONTROL 7.5.2

Respiratory Protection

To the extent practical, reliance for safety should be placed on engineered controls rather than on the use of administrative controls or personal protective equipment. However, there may be situations that require the use of individual respiratory protective equipment. Caution and judgment must be exercised in determining the need for respiratory equipment because it may increase physical stress, impede the worker's vision and extend the time required to complete the work. Thus, the use of respiratory equipment may lead to increased risk of injury and increased external radiation exposure (Dooley and Barresi, 1994; Lee, 1994). The goal of the protection program is to control the total radiation dose and to limit the total risk to the individual. When respiratory protection is required, a n adequate program should include (ANSI, 1984; 1992; Colton et al., 1991): 1. 2. 3. 4. 5. 6.

7. 8.

medical evaluation and surveillance specification of respirator type testing the fit on individuals (individual fit testing) user training maintenance and cleaning testing the effectiveness of the respirator a bioassay program evaluation of other risks, e.g., heat stress, falls, trips, etc.

The need for respiratory protective equipment can be reduced by such measures as replacing dry operations with wet operations or confining the radioactive material in gloved enclosures or chemical hoods. 7.5.3

Protective Clothing

Protective clothing is provided to workers to prevent contamination of their skin and personal clothing. The protective clothing should be durable and effective. Tightly woven cotton or synthetic fabrics are usually acceptable for work in dry areas (in some cases disposable paper clothing may be acceptable). Plastic or rubberized garments are necessary when liquids are involved. The effectiveness of protective clothing depends on: 1. the type of garment 2. selecting the appropriate clothing for the job

7.6

RECORDS 1 71

3. training in the proper techniques for donning and removing the clothing 4. proper laundering The required protective clothing should be specified by the health physicist and the requirements should be recorded on a RWP or procedure (Section 7.2.2). Usually coveralls, gloves and shoe covers will be sufficient to limit the probability of contamination. More elaborate protective clothing programs are needed when the potential exists for serious contamination and internal exposure. Outer garments should always be considered to be potentially contaminated. Because protective clothing can become damaged with use, it should be inspected periodically to verify its continued integrity and usefulness. 7.6 Records

Internal radiation dose records should be maintained to demonstrate compliance with dose limits and administrative dose guidelines and to assist in the evaluation of the effectiveness of the internal dose control program. Recommendations for the content of occupational dose records are given in NCRP Report No. 114 (NCRP, 1992) and include the following: the annual intakes of radioactive material radionuclides deposited in the body exposure routes physical and chemical forms of the radionuclides purpose of bioassay measurement (i.e., baseline, termination, diagnostic or routine) date and location of suspected intake type of bioassay measurement a listing of the bioassay data used in the determination of the equivalent dose information to enable linkage to procedures, calibration factors, geometry, background and energy resolution checks, and confidence levels metabolic and dosimetry models used assumptions used to estimate intake of radioactive material and the committed effective dose magnitude and location of the deposition of specific radionuclides

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13. identification of the individuals who were involved in making the estimates of intake, deposition and committed effective dose 14. organ or tissue equivalent doses and committed effective dose

In addition to the records maintained for the personal dosimetry program, records should be maintained of the s w e y s of airborne radionuclides that are performed including the information that is detailed in Section 7.4.1

8. Control of Low-Level

Radioactive Waste Radioactive waste is unwanted material that contains, or is contaminated with, radionuclides. In general the management, transportation and disposition of hazardous wastes, including radioactive wastes, are matters of great public concern and are strictly regulated by federal and state agencies. When designing a program for control of radioactive wastes it is the responsibility of the facility management to become thoroughly familiar with currently applicable regulatory requirements and community concerns. Based on these external constraints, standards and control levels (e.g., release criteria) appropriate to the facility must be defined that are consistent with applicable technical considerations, and facility-specific factors. This Section addresses general technical aspects of the control of low-level radioactive wastes as part of a radiation safety program. It is not intended to be directly applicable to facilities, such a s nuclear power plants, that generate large quantities of radioactive wastes and that have separate waste management programs. Also excluded is specific discussion of the management of high-level waste (e.g., spent nuclear fuel and some fuel reprocessing wastes) and transuranic wastes. I t should be noted that the regulatory definitions of waste categories generally are not based solely on the radiation hazard of the materials. Thus, for example, some wastes defined as "low-level" for regulatory purposes, may be just a s rahoactive as some wastes defined as "high-level," and such "low-level" wastes are also excluded from consideration in this Section. Radioactive wastes may be generated by the primary activity of the facility (e.g., animal carcasses from research, activated components from accelerator operation, contaminated tools from reactor facility maintenance) or by the radiation safety program itself (e.g., contaminated protective clothing or survey instruments). As discussed in this Section, mixed wastes are those that contain both radionuclides and other hazardous substances (e.g., organic solvents, asbestos, heavy metals, or reactive chemicals). The reader is

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referred to regulatory definitions to establish whether or not a specific mixture must be considered a "mixed waste* for regulatory purposes. 8.1 Minimizing the Production of Waste Facilities that use, manufacture, or in some way generate radioactive materials should establish a program for minimizing the generation of waste that is, or could be, radioactive. The program should include procedures that separate materials that may become radioactive from other potentially hazardous materials to avoid the generation of mixed wastes. Each proposed new practice should be reviewed for its waste generation potential. Designs, materials, processes and procedures that reduce the potential quantity of waste produced and avoid the generation of mixed waste should be adopted whenever it is practicable and cost effective. The following principles should be considered for waste minimization: 1. waste generation should be avoided or reduced a t the source, wherever feasible 2. unavoidable wastes should be recycled in a n environmentally safe manner, whenever possible 3. unavoidable, nonrecyclable waste should be reduced in volume and treated, whenever feasible, to render it less hazardous, toxic and harmful to the environment 4, land disposal of waste should be conducted in an environmentally safe manner 5. waste minimization is not achieved merely by diverting waste from one medium to another (e.g., land to waterway) 8.1.1

Practices for Minimizing Waste

The amount of radioactive waste generated that must be disposed of can be minimized by incorporating certain practices into the design and operation of the facility. The facility should be designed to minimize the number and the size of the areas where potential activation or contamination is possible. Storage areas for nonradioactive supplies and parts should be located outside areas in which these materials might become activated or contaminated. Where activation is a potential concern, construction materials

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should be chosen to minimize the generation of unwanted radioactive material. Training programs for workers a t the facility should include instruction in waste minimization principles and techniques. Policies should be established to restrict materials from being brought into areas where they may become activated or contaminated. Packaging material, for example, should be removed and disposed of in clean areas. An assortment of tools and equipment should be reserved solely for use in areas where activation or contamination is possible. These reserved tools and equipment should be identified, conspicuously marked, issued and used under strict controls. Equipment should be decontaminated, if necessary, and reused. Recyclable or burnable materials should be used rather than items that require land burial. Disposable materials (e.g.,for protective coverings and clothing) should be selected to be compatible with waste-processing systems, volume reduction and waste disposal requirements. Potentially activated or contaminated material should be segregated from nonradioactive or uncontaminated material by providing separate containers for nonradioactive and radioactive trash, or by surveying individual items in unsegregated trash. Separate containers can also be used to segregate reusable radioactive items such as tools, instruments, protective clothing and respirators from disposable radioactive wastes. This minimizes the effort that is needed to sort out the reusables before disposing of the trash. It may also be feasible to segregate wastes by half-life. Wastes containing only short-lived radionuclides can be stored for decay to levels that permit the material to be disposed of as nonradioactive waste (see Section 8.5). Background information on waste minimization is presented in documents of the NRC (1994) and EPA (1993).

8.1.2

Practices for Reducing Mixed Waste

Generation of mixed waste can be avoided by establishing practices that restrict the quantities of hazardous materials (e-g., paints, solvents, chemicals, cleaners and fuels) that enter areas where they might become contaminated or activated. When hazardous materials are needed in such areas, appropriate measures should be instituted to prevent or minimize their activation or radioactive contamination. In processing operations, hazardous materials should be reduced or eliminated whenever possible through changes in the

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process and substitution of materials. When mixed wastes are unavoidable, they should not be combined with other wastes. Procedures should be established to identify those materials that are suspected of being mixed waste and to segregate them from materials that are either hazardous or radioactive only. Such practices can reduce the quantity of material that must be considered mixed waste.

8.2 Decontamination and Reuse of Tools and Equipment Tools and equipment such as survey instruments or laboratory glassware that have been used in areas in which they may have become radioactive or contaminated should be surveyed for radioactivity after use. Those with removable contamination should be decontaminated prior to removal from the restricted area. It is good practice to decontaminate tools and equipment before they are reused or stored even if they remain in the same restricted area. However, if complete decontamination is not possible, a n assortment of tools and equipment can be reserved solely for use in areas where activation or contamination is possible. These items must be conspicuously marked to reduce the possibility that they might be moved to a clean area. Periodically an inventory of radioactive tools and equipment should be taken. Any discrepancies in the inventory over time should be investigated and resolved. Methods that are used to decontaminate tools and equipment include dry wiping, rinsing with water, washing with water and detergent, treatment with special decontamination solutions, ultrasound treatment, and abrasive decontamination. Selecting the method to be used requires evaluating the degree and type of contamination, the materials and construction of the contaminated item, and the criteria for acceptable levels of residual contamination. In selecting a decontamination method, consideration should also be given to the potential radiation exposures of all workers involved, including those performing the decontamination, and the potential for generating additional radioactive or mixed wastes.

8.3 Collecting, Sorting and Classifying Waste Conspicuously marked or colored bags, metal drums or cans lined with plastic bags should be used to collect radioactive solid wastes. Depending on the materials present and the quantities

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expected, separate and easily distinguishable containers should be provided for different classes of waste, e.g., nonradioactive, potentially radioactive or contaminated, definitely radioactive or contaminated, reusables and recyclables. All but the nonradioactive trash containers should bear "radioactive materials" labels. Contaminated waste should be clearly marked. Sharp objects should be specially packaged before collection to prevent punctures of the containers. Before materials are packaged for disposal as radioactive waste, they should be checked for nonradioactive and reusable items. Radiation doses to workers who sort waste can be controlled better if waste material is segregated and packed in clear plastic where it is generated. Automated devices that detect surface or volume radioactivity may be useful for confirming the status of material that has been segregated as nonradioactive. Tests of these devices should be conducted prior to their being used to ensure that radioactivity would be detected a t the desired level of detectability. Frequent checks should be made of the operation and detection sensitivity of the system. Low-level wastes that are wet, flammable or corrosive should be dried and neutralized before they are placed in containers for disposal. Spray cans should be vented or punctured to ensure that they will not explode when compacted. Wastes that are nonradioactive but that contain materials generally associated with radiation control, e.g., yellow plastics, protective clothing, or items that are marked with radiation symbols or other radiation control indicators should be shredded or defaced before they are disposed as nonradioactive waste. 8.4 Radioactive Waste Volume Reduction

Solid waste volume reduction techniques should be adopted to minimize the space needed for storage and disposal of radioactive waste. These techniques include compaction and incineration. Compaction can substantially reduce the volume of solid waste that is destined for shipment for disposal off-site. Where large numbers of scintillation vials are used, wastes can be processed by crushing the vials. The radioactive scintillation fluids can then be separated from the fragments of the vials which can be cleaned and disposed as nonradioactive waste. Material containing short half-life radionuclides can be stored for decay and then disposed as nonradioactive waste.

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8. CONTROL O F LOW-LEVEL RADIOACTIVE WASTE

Incineration is an effective volume reduction method for putrescible materials such as animal carcasses. It is also useful for disposing oils and fluids that are used in liquid scintillation counting. Incineration can also be used for volume reduction of combustible trash. If incineration results in radioactive ash, the ash must be disposed as radioactive solid waste, or if it also contains hazardous materials, as mixed waste. Similarly, if incinerator off-gases are filtered, contaminated filters must be disposed as solid radioactive waste. 8.5 Storage of Waste Radioactive animal carcasses should be stored frozen while awaiting disposal. If space permits, they should be stored long enough to allow any short half-lived radionuclides to decay away. If only short half-lived radionuclides are present in the waste material, it may be feasible to store the waste in containers for a long enough time to allow the activity to decay to a level a t which the material may be disposed as nonradioactive waste. Radioactive or mixed wastes that are not stored for radioactive decay should be prepared and packaged for disposal as soon as collection and sorting are completed. They may then be stored in those packages until they are shipped for disposal. Radioactive materials that are stored for decay or packaged for disposal should be protected sufficiently to prevent deterioration of the containers and to minimize the likelihood of the spread of contamination. Storage locations should be clearly identified as radioactive material areas and periodically inspected. To the extent possible, wastes should be stored inside soundly constructed buildings. To provide security and reduce radiation exposure, the storage area should be isolated from the rest of the facility and access to the area should be restricted. 8.6 Disposal of Waste Many factors must be considered in the selection of appropriate waste disposal methods, including technical feasibility, cost, regulatory requirements, liability, and public and environmental impacts. Disposal methods include on-site incineration or burial and off-site disposal. Incineration primarily burns the nonradioactive constituents of the waste and is best thought of as

a volume reduction technique, because the resulting radioactive ash requires disposal as radioactive waste. Although on-site burial may be considered a technically feasible means for disposing some types of slightly radioactive wastes including animal carcasses, regulatory requirements and public concerns oRen make such an approach problematic. Among the items that must be considered when establishing a burial site are the nature of the waste and the site, possible transport of the radionuclides in the soil and groundwater, and the depth of the burial site relative to the groundwater table. Radioactive wastes that will be shipped for disposal off-site should be packaged, labeled and shipped in a manner that ensures the integrity and the identity of the package and meets the acceptance criteria of the receiving organization. Mixed wastes are generally packaged for off-site treatment and disposal by specialized contractors. Incineration is particularly effective for oil and may be a viable disposal option for other waste depending on its composition and the quantity of radioactive material, the operating parameters of the incinerator and the type of ash produced.

8.7 Recycling of Waste Recycling of certain radioactive materials can be better for the environment than their disposal as waste. Therefore, slightly contaminated or activated materials (e.g., concrete and scrap metals) should be evaluated for possible reuse. For example, concrete scrap from demolition might be used as fill for erosion control; metals could be re-fabricated for use in structures like bridges.

8.8 Records

Records of the disposal of radioactive and mixed wastes should be maintained in a retrievable and legible form for the lifetime of the facility involved. Guidance for this record keeping is found in NCRP Report No. 114 (NCRP, 1992).

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8.9 Recommended Additional Reading Evaluation of Abrasive Grit-Hgh Pressure Water Decontamination, Westinghouse Electric Corporation EPRI-NP-2691, Electric Power Research Institute, Palo Alto, California, 1982. Guidelines for Radiological Protection at Nuclear Power Stations, INPO 88-010, Institute for Nuclear Power Operations, New York, 1988. Plant Decontamination Methods Review, EPRI-NP-1168, Electric Power Research Institute, Palo Alto, California, 1981. Radioactive Waste Management, Berlin, R.E. and Stanton, C., Wiley, New York, 1989. Radioactive Waste llchnology, Moghissi, A.A., Godbez, H.W. and Hobarts, S.A., American Society of Mechanical Engineers, New York, 1986. Radiological Control Manual, DOE EH-0256T, U.S. Department of Energy, Washington, 1992. Radiological Protection, 3rd ed., Shapiro, J., Harvard University Press, Cambridge, Massachusetts, 1990. Radiological Protection at Research Reactor Facilities, American National Standards Institute/American Nuclear Society 15.11-1993, American Nuclear Society, La Grange Park, Illinois, 1993. Volume Reduction of Low Level Radioactive Waste, American National Standards Institute/American Nuclear Society, 40.35-1991, American Nuclear Society, LaGrange Park, Illinois, 1991.

9. Control of Exposure to the Public The NCRP has recommended limits for exposure of members of the public to sources of ionizing radiation produced or enhanced by human activities. Other limitations on public exposure may be specified in the license under which a facility operates or by other legal authorities. Important goals for the radiation safety staff are to ensure that public exposures due to operation of the facility meet these exposure limits and, hrther, that public exposures are ALARA. These goals are achieved by controlling radiation fields emanating from the facility and by controlling releases from the facility to the environment. Control begins with proper facility design and includes appropriate procedures for operation within the design envelope. Assurance that appropriate levels of control have been achieved is frequently provided by programs of effluent and environmental monitoring. The size and complexity of the monitoring program that will be required depend upon the types and quantities of radionuclides present in the facility and the radiation fields that its operation generates.

9.1 Standards and Guidance

The NCRP recommends that continuous exposure of members of the public be limited to an annual effective dose of 1 mSv. For individuals exposed infrequently, a n annual effective dose limit of 5 mSv is recommended (NCRP, 1993b). These limits exclude exposures from natural background radiation and radiation exposure associated with medical diagnosis and treatment. The total dose from both internal and external exposure must be less than the stated limits and should be kept ALARA. It should be noted that the limits apply to the sum of all exposures, not to each source individually.

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Further limitations on radiation exposure of members of the public may be dictated by the regulatory agencies that license operation of the facility or regulate its operations. Specific environmental protection regulations may also apply.

9.2 Control o f Off-Site Exposures

Radiation fields emanating from the facility are controlled by appropriate shielding of components or equipment that are sources of radiation. The choices of control measures are highly dependent on the nature of the facility and its processes, the quantities and types of radionuclides employed or processed, and the levels and types of radiation produced. The facility management must ensure that techniques for control of releases of radioactive materials are adequate and that they are functioning at a satisfactory level. As discussed in Section 9.2.1, this demonstration may not require extensive monitoring. However, a n operating facility will have special monitoring requirements that depend upon the mixture of radionuclides released and the radiation fields generated during operation. There are aspects of effluent monitoring that are common to a wide variety of facilities and these are discussed in Sections 9.2.2 through 9.2.4. The radionuclide measurement capabilities required for effluent monitoring depend on the composition of the effluent streams. The term "monitoring" is here defined in the broadest sense. It includes assessment of releases by collection of periodic integrated samples whose content may be determined by subsequent laboratory analysis. Gross activity measurements may be sufficient if the effluent activity is very much less than established limits. In general, however, gross activity measurements are discouraged and techniques that provide information about specific nuclides are preferred. Periodic determinations of specific isotopic composition may suffice if it is known that the radionuclides in the effluent stream are consistently present in relatively constant proportions. Low-energy beta emitters and weak gamma emitters may require special monitoring methods to assess releases. The following sections are not intended to constitute a manual for the design of effluent monitoring systems. However, some of the important considerations related to systems for monitoring airborne, liquid and solid effluents are discussed.

9.2CONTROL OF OFF-SITEEXPOSURES /

9.2.1

83

Determining the Need for Monitoring

The need for, and extent of a n effluent or environmental monitoring program should be determined by both an initial assessment of the types, quantities, and forms of radioactive materials that are expected to be present in the facility and a review of the anticipated containment, processing, and disposition of these materials. If the quantities of materials are relatively small, if the containment is good, and if the distribution of the materials within and through all processes and activities are well known and adequately documented, it may be possible to demonstrate that the maximum potential release to the environs and potential doses to nearby residents would be too small to justify an effluent or environmental monitoring effort. Screening models are available (NCRP, 1996) to provide guidance in the determination of the need for monitoring. The input to these models consists of data on the types, quantities, and forms of the individual radionuclides present in the facility, on locations and modes of potential releases, and on the characteristics of the surrounding environment and population. In the application of the screening models, the degree of complexity may be increased incrementally to provide greater confidence in the results, until a conclusion with a n acceptably small uncertainty is reached. Before planning or acquiring a monitoring system, the required accuracy and precision of the characterization of the radionuclides in each effluent or waste stream should be determined. Detailed knowledge and documentation of inventories and processes may often be used in conjunction with partial monitoring to reduce operating costs. If the distribution and movement of radionuclides through the entire process or facility is thoroughly documented, the quantities released to waste streams or to the environs might also be ascertained with adequate accuracy. In performing mass balance or throughput calculations, the uncertainties in each step must be evaluated and combined to provide a reliable estimate of the overall uncertainty in each effluent or waste product that is subject to specific limits for release or disposal. For example, estimates of the upper limits of radioactive materials in solid wastes, based on process knowledge with small uncertainties, may be adequate for purposes of transportation and disposal. If process knowledge is relied upon in lieu of routine monitoring, special attention must be paid to the monitoring that might be needed in the event of accidental release of radioactive material. Systems installed for routine monitoring are typically designed to also provide emergency monitoring capabilities. However, if no

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routine monitoring system is planned, emergency monitoring must be considered a s a n independent requirement. 9.2.2

Monitoring Airborne Efiuents

When establishing a monitoring program for airborne effluents, it is first essential to understand the ventilation system for the facility and to define the point(s) where discharges to the environment can occur. It is also important to determine the potential for routine and accidental releases from each discharge point. This determination requires knowledge of the process, the potential for abnormal conditions, and the movement of air through the ventilation and effluent treatment systems. Effluent monitoring requirements and criteria should be considered during the design of the ventilation system. Knowledge of the air flow rates a t the points of release is essential to assess the quantities of radionuclides that are released. Because flows may vary with operating conditions and with time (e.g., due to changes in blower capacity and filter loading), routine records of fan operations and periodic measurements of discharge flow rates are important components of the airborne effluent monitoring program. Airborne effluents are normally monitored at a point that is downstream of all effluent treatment systems and of the exhaust fans, just prior to discharge to the environment. The primary consideration in selecting a monitoring location is the requirement that the air withdrawn from the effluent stream contains a representative sample of that stream. Consensus guidance for selection of monitoring locations has been prepared (ANSI, 1969); however, specific governmental regulations that differ from the ANSI guidance may apply. The ANSI guidance addresses questions such as the need for isokinetic sampling of particles and the desired separation of the monitoring point from disturbances of the airflow in the exhaust duct or stack. For large ducts or stacks, guidance is provided regarding the number of sample withdrawal points needed to assure that the composite will represent the average concentration being discharged. In exhaust systems that have several ducts that are combined in a plenum, uniformity of flow across the exhaust duct is not a suficient condition for a uniform distribution of contaminants in the discharge. The presence of the exhaust blower upstream of the sampling location will help to assure that the radionuclides are well mixed in the exhaust but it is not a guarantee. A reliable evaluation of the

9.2 CONTROL OF OFF-SITEEXPOSURES / 85

degree of mixing in the exhaust gases of contaminants released from various parts of the facility can be performed using nonreactive tracer gases (e.g., He and SF6). Measurements of concentration profiles in the exhaust duct that reflect constant releases from several important process areas within the building will show whether multiple probes are necessary to ensure that a sample that is representative of the effluent is collected. Evaluations of the requirements for representative sampling for normal operating conditions and those expected following a n accident, when ventilation system flows may be quite different, are both important. Transport of the sample from the monitoring location to the point of collection is another factor that may affect the choice of monitoring location. Short sampling lines are generally recommended to reduce the potential that material may be lost because of deposition or impaction on the wall of the line. It may be necessary to compromise between competing monitoring goals. In any case, it is important that any bias due to loss of material in transport be understood and accounted for when monitoring results are reported. An experimental determination using the installed Sampling line presents challenges but can be performed (Curtis and Guest, 1986). Modeling approaches to this problem suffer from the difficulty of having to extrapolate from relatively short-term laboratory test conditions to long-term use of installed sampling lines. Models of transport of radioiodines have been developed and parameters measured experimentally (Unrein et al., 1985). If the effluent contains radionuclides in particulate form, it is important to collect information about the particle size distribution. Models for estimating losses of particles during sample transport, as a function of size and other parameters, are available and can be used to select sampling line parameters that will minimize deposition and impaction losses (Anand et al., 1993). If the sample collected from the effluent is not or can not be analyzed in real time, a sampling frequency must be established. The selected frequency will depend on the variability of the radionuclide release rate, the chance that unplanned releases will occur, the likely magnitude of any unplanned releases, and other factors. In some situations, a separate monitoring system may be necessary for post-accident conditions because the demands for information about routine effluents and about accidental releases are not achievable in a single system.

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9. CONTROL OF EXPOSURE TO THE PUBLIC

Monitoring Liquzd Effluents

The first requirement for developing a monitoring program for liquid effluents is an understanding of the source of radioactivity, the processes that affect it, and the effluent piping and treatment systems. For liquid waste systems, there may be multiple streams with different characteristics, waste generation rates, and treatment requirements prior to discharge. Releases of some liquid wastes may be continuous, but collection of liquid wastes in a tank prior to batch discharge is also common. Routine determination of the liquid flow rates a t continuous release points and of volumes of liquid wastes that are discharged in batches is essential to estimation of the total radionuclide discharges. Waste water that has been stored prior to batch discharge can be sampled and analyzed for a broad range of contaminants. This will frequently provide greater sensitivity than on-line monitoring. Collection of representative samples of the liquids that are discharged can be challenging because some radionuclides may be in solution while others may be present in suspended solids or as sediments in the tank, with the proportions dependent upon variables such as pH and temperature. Prior to grab sampling, the liquid in the tank should be thoroughly mixed. If the tank can not be mixed, the inlet stream should be sampled proportionally or monitored throughout the time that the tank is being filled. It may also be desirable to obtain multiple samples from the waste transfer or discharge line to confirm release estimates based on sampling of a tank or the inlet flow. If liquid waste discharges occur continuously but flows are variable, proportional sampling of the effluent is recommended. Monitoring may be accomplished using instrumentation that provides data in real time. While this technique offers the possibility of prompt notification and automatic shut-off capability if elevated releases are detected, the method is not uniformly applicable. 9.2.4

Monitorirg Solid Waste

Shipment of solid wastes from a facility is not a major source of public exposure. It is, however, necessary to document the quantities of radionuclides present in such wastes and to assure that the external exposure rates from shipping containers are a t acceptable levels. Records of the weight and volume of waste containers, as well as quantification of the radionuclide content to be shipped are a n integral part of the documentation that is needed, but many

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additional requirements may be specified by federal or state regulatory agencies. There are special requirements for estimating the amounts of long-lived radionuclides and fissile materials present in the wastes. If the radionuclides in waste material are energetic gamma emitters, it is possible to use external measurement techniques to assess the activities of those radionuclides in the waste container. In addition, techniques for assaying the fissile material content of waste containers have been developed. However, if the waste is contaminated primarily with low levels of pure alpha or beta emitters, sampling of the waste will likely be necessary. The amount of effort devoted to characterization of such wastes should be commensurate with the level of radioactivity and applicable regulations. Paramount to estimation of the content of such radionuclides in the waste is the collection of a representative sample. The sampling technique of choice will depend on the form of the waste material. Coring has been used to obtain samples of wastes in semisolid form prior to solidification. For heterogeneous wastes, the component waste streams may be analyzed prior to consolidation or composites of grab samples may be used. It may also be possible to collect representative core samples of heterogenous wastes following compaction. Initially, an assessment of the variability from sample to sample should be conducted to provide guidance for routine sampling of such waste materials. 9.3 Environmental Monitoring

Environmental monitoring serves a number of purposes. First, it provides a mechanism to assess the effluent monitoring program and in many situations can be used to determine whether environmental concentrations are consistent with the release rates that were estimated. Even if environmental concentrations are extremely low, as is often the case, the data can be used to place upper limits on the magnitude of releases that occurred. A second use of environmental monitoring data is to assess the level of exposure of members of the public due to releases or external radiation fields from the facility. Again, very low radioactivity concentrations or radiation doses serve at least to estimate the maximum exposure of members of the public from radionuclides released by the facility and from external radiation fields. A third use of environmental monitoring data is to identify environmental mechanisms that lead to elevated concentrations of radionuclides in biota or other materials and to quantify the

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magnitudes of the associated concentration factors. The analysis of particular environmental pathways will aid in identification of the critical group of persons whose diet or habits cause them to be most highly exposed to facility effluents. Finally, a sound environmental monitoring program can support and improve the credibility of estimates of public exposure that are made by facility staff. Alternatively, a poorly designed program that lacks adequate sensitivity, quality assurance, or other important components can destroy the credibility of such estimates. The type of environmental monitoring program that is appropriate for a particular facility depends on a number of factors. These include the direct radiation levels produced around the facility, quantities of radionuclides released from the facility, the isotopic composition, chemical and physical forms of effluents, and points of release to the environment. Also of interest are the expeckd behavior of the radionuclides in the atmosphere and terrestrial or aquatic environment to which they are discharged and the ways in which that behavior may be modified by particular features in the environs of the facility. Land and water uses comprise a third set of factors: the nearby population distribution, its habits, and source of drinking water; local agricultural production and gardening activities; and local recreational, subsistence, and commercial harvesting of fish or other aquatic organisms. Facilities that use or process small quantities of radionuclides may be able to employ conservative calculations to demonstrate compliance with regulatory requirements. One technique for accomplishing this is the use of screening models for atmospheric releases, liquid releases, and land disposal (NCRP, 1996). Other facilities, which handle larger quantities or more hazardous materials, routinely release radionuclides to the environment, or have the potential for large accidental releases, will find a n environmental monitoring program to be beneficial for one or more of the reasons set forth above. The following subsections do not comprise a manual for the design of a comprehensive environmental monitoring program, but are intended to assist staff members at the latter class of facilities. 9.3.1

Preoperational Monitoring

Conduct of an environmental monitoring program prior to facility start-up allows for testing of the monitoring techniques and analytical procedures that are needed to measure radionuclides in

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the environment. It also allows the facility operator to document preoperational environmental radionuclide levels and their variability. This latter goal is less important now than in previous years when global fallout of radionuclides from weapons testing made cyclic contributions to local ambient levels. However, remnants of the long-lived radionuclides released during the above ground weapons testing era are still found in soils, sediments, and to some extent in other media. Occasionally, these radionuclides may be significant. Some investigation of other facilities near the site is also warranted. I t may be known that hospitals or other institutions release radionuclides into a waterway shared by the new facility. The importance of such releases and of residual fallout radionuclides to the environmental monitoring program will depend on the type of facility and the radionuclides produced or handled within it. The preoperational period is also the time for investigation of the factors cited above that are important to the doses that will be received by nearby populations. Careful review of mechanisms of radionuclide releases from the facility and investigation of the behavior of the radionuclides that are expected to be released will identify the most important radionuclides and environmental exposure pathways that lead to exposure of the critical groups. This information, together with results of meteorological and hydrological investigations, will aid the design of the environmental survey and in the selection of environmental monitoring locations. 9.3.2

Operational Monitoring

The operational monitoring program will provide data that can be used to estimate, or a t least provide satisfactory bounds for, the dose received by the critical population groups. The period over which the dose is averaged is likely to be one year, but specific licensing requirements may identify a shorter period. Although relatively long-term averages are used for dose assessment, environmental sampling frequency also depends on the radionuclides expected to be released, their half-lives, and their behavior in the environment. Identification of critical radionuclides and pathways, including external exposure, in the preoperational investigation will allow the monitoring program to focus on those modes of public exposure that are most important. However, the program should also have the capability to identlfy unusual occurrences, unanticipated

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radionuclide releases, build-up of radionuclides in the environment and unexpected environmental pathways. Coordination of the monitoring program with the dose assessment procedure will assure that the monitoring data collected include those desired for dose assessment. In general, it is desirable to monitor exposure along a critical pathway a t the point closest to the exposed person. For example, in the air-grasscow-milk-child pathway, measurements of radionuclides in milk will be more reliable indicators of the dose to the child than measurements of air or pasture grass contamination. When media or organisms that concentrate radionuclides are not themselves consumed by humans, but are used as indicators for concentrations in a human food item, the relationship between the indicator and the foodstuff must be defined in preoperational or early operational surveys. Measurements of such correlations in the local environment are preferred over those obtained from the literature unless it is known that the two environmental settings are very similar. Even then, confirmatory measurements are desirable. The operational environmental monitoring program will also provide a check on the effluent monitoring program. Periodic review of the two data sets and relevant dispersion estimates is required to determine whether measured environmental levels are consistent with those expected from effluent measurements and expected levels of dispersion and food chain concentration. Monitoring locations that are selected on the basis of sitespecific average meteorological conditions may not provide the desired information for a short term accidental release to the atmosphere. However, routine monitoring operations help to build capabilities that are needed for emergency monitoring. Operational systems may provide direct exposure rate data or measures of integrated external doses and concentrations a t established air and water sampling locations. These data will help to document the impact of an abnormal condition on the local environment.

9.4 Measurement Methods

The physics underlying radiation measurement techniques and a number of radioactivity measurement procedures are described in NCRP Report No. 58 (NCRP, 198513). The techniques presented in that report can be applied for analyses of samples from both effluent and environmental radioactivity monitoring programs.

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Measurements of environmental radiation fields and environmental sample collection, preparation and analysis procedures are discussed in NCRP Report No. 50 (NCRP, 1976b). References to many radiochemical procedures are included as are discussions of common laboratory instrumentation. The manual of procedures of the Environmental Measurements Laboratory (EML), which is periodically updated, is an excellent source of information on techniques of radiochemical analysis, sample collection, field measurements of radiation fields, and radioactivity measurements (EML, 1990). The literature on radioactivity measurement techniques, procedures for radionuclide sampling and analysis, and solution of special problems is voluminous. Selection of appropriate sampling, analysis and measurement techniques for effluents and for environmental samples will depend on the radionuclides that are released, their concentrations, the nature of the site and its proximity to residential and farming areas, and other factors. The three references given are intended to serve as general guides and starting points into the investigations into methods suitable to the needs of a particular facility. 9.5 Dose Assessment Among the goals of the effluent and environmental monitoring programs is the assessment of radiation doses received by members of the public in the vicinity of the facility. The complexity of the dose assessment procedure depends on the type of facility, the amounts and nature of the radionuclides released from it and the nature of any direct radiation fields generated during operations. For facilities handling small amounts of radioactive materials, screening level assessments may suffice to show compliance with good radiation protection standards. In NCRP Report No. 123 (NCRP, 1996), the NCRP has provided generally conservative screening models for environmental releases [see also NCRP Commentary Nos. 3 and 8 (NCRP, 1989d; 1993c)l. These off-the-shelf models, with conservative parameters, may be satisfactory to determine upper bounds for exposure of members of the public. Other facilities may require more complex modeling using parameters that are more representative of actual site conditions. Data to support the choices of parameters can be obtained from site-specific environmental studies or from the environmental monitoring program. Facilities that require more complex modeling may have on-site meteorological and hydrological

92 / 9. CONTROL OF EXPOSURE TO THE PUBLIC

monitoring programs that provide data for estimating dispersion of effluents. Whether the dose assessment is performed using a simple screening model or a complex, site-specific model, it is essential that the assumptions and results be thoroughly documented. Particular aspects of the calculation that should be included or referenced in the documentation are: 1. the effluent release data, the sampling and measurement techniques used to obtain them, and relevant instrument calibration and quality assurance programs 2. the basis for the estimated dispersion of atmospheric or liquid waste releases prior to reaching a point of public exposure 3. the environmental pathways considered, the specific parameter values employed in the calculations (e.g., deposition velocity or concentration factors), and the basis for their selection 4. assumptions made about the consumption of water and agricultural products, e.g., vegetables, meat, fish and shellfish 5. assumptions made about occupancy and particular activities that may be important for the critical group, e.g., fishing, boating, swimming 6. the sources of dose conversion factors employed in the calculations of doses from intakes of specific radionuclides and in calculations of doses due to exposures to contaminated environmental media 9.6 Quality Assurance

As noted in Section 3.6, the results of measurements made as part of a radiation safety program must be reliable. This is particularly true for the measurements of radionuclides in effluents and in environmental media discussed above. A facility operator should assure that all steps in the collection, analysis and reporting of effluent and environmental samples are performed in an appropriate way and that the use of the results in a dose assessment procedure is consistent with the nature of the information. There are a number of steps that are needed to provide confidence that the effluent and environmental monitoring data are reliable. All aspects of the program should be carried out by

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qualified individuals using appropriate procedures that have been reviewed and adopted to meet the goals of the program. The instrumentation used for measurements of radiation and radioactivity must be of adequate quality and calibrated using reference standards prepared by the National Institute of Standards and Technology (NIST) or other qualified sources. Routine checks of the instrumentation, analysis of duplicate or replicate samples, and comparisons with independent laboratories of the results of analyses of split samples are quality-control activities that contribute to data reliability. Quality assurance activities should be carefully recorded and written reports of the work should be prepared routinely. The quality-control activities that are performed by the facility operator will depend upon whether the effluent sampling and analysis steps are performed in-house or by a contractor. However, overall responsibility for quality assurance of the monitoring data rests with the facility operator. Some aspects of the quality assurance program may be dictated by the license for the facility or by guidance issued by the licensing or other governmental authority. 9.7 Records

Documentation of the measurements of effluents from the facility is an essential component of the monitoring program. This documentation should include the data collected, information about unusual occurrences (e.g., sampler malfunction) and assumptions made when estimating unmeasured releases. In addition, it is equally important to maintain records of effluent flow rate measurements, instrument calibrations, analytical techniques employed and the modifications of those techniques with time, and quality assurance procedures for the various components of the measurement program. The environmental monitoring program must be similarly well documented. In addition to documentation of measurement techniques, as discussed above, the meteorological and hydrological measurement programs must also be documented, including records of equipment calibration and testing and appropriate quality assurance measures that were employed in data collection and analysis. The capability for long-term safe retention of these records must be developed and maintained. Further recommendations regarding such records are provided in NCRP Report No. 114 (NCRP, 1992).

10. Radiation Safety Instrumentation The instruments that are needed for a radiation safety program include those that: 1. are required to perform surveys of areas where there is a potential for external exposure to radiation fields 2. are required to detect radioactive contamination on surfaces

3. monitor the ambient radiation and airborne radionuclides in the workplace 4. monitor the external radiation environment as well as the

radionuclides that are released to the environment outside the control of the facility

5. are used to document the radiation exposure of individuals who work in the facility Instruments that are chosen for surveys and personal, area or environmental monitors for external radiation must be capable of measuring both the primary and scattered external radiation fields that are present in and around the facility. Instruments that are chosen for measurement of radionuclides in the air or in other effluents must be capable of detecting all the radionuclides that are used within the facility a t levels that are at, or below, administrative exposure guidelines or reference levels. This instrumentation must be appropriately calibrated and maintained, and proper records must be kept of the instrumentation types, use, calibrations and maintenance. This Section describes the specification, calibration, maintenance, use and record-keeping requirements for the instruments that are used to implement a radiation safety program.

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10.1 Instrument Specification When the performance specifications and characteristics of an instrument are being developed, the following factors should be considered (ANSI,1989b; 1989~): 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11.

accuracy precision sensitivity power supply stability, lifetime, type environmental factors such as temperature, pressure and humidity range and type of readout (analogldigital) rate or integral measurements energy and angular dependence computer interface requirements weight, portability, durability interference from electromagnetic fields

The most important factors for instruments that will be used for radiation measurements are the types of radiation that are to be measured and the conditions under which the measurements will be made. These factors will determine the type and performance specifications of the instrument that is most appropriate for the measurement. Performance specifications should address the accuracy and precision that are required, the type, stability and lifetime of the power supply, the behavior of the instrument under extremes of temperature and pressure, the dynamic range, sensitivity, and angular and energy dependence of the instrument. Additional specifications may include the type of readout and readout scale, whether the instrument measures dose (or exposure) rate or integrates dose (exposure), or both, and whether a computer interface is needed. Requirements may be written that specify portability or weight and instrument durability. Certain other specificationsmay cover ease of decontamination, protection against moisture, protection against the effects of electric and magnetic fields, as well as radiofrequency and microwave radiation. The provision of features such as audible alarms and status indicators may also be specified. Specifications for instruments that are to be used to count radioactive samples or for spectrometry should include limits on dead time and energy resolution. Carefully written specifications can ensure that the proper instrument will be purchased for the radiation measurements that need to be made.

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Acceptance of any radiation measuring instrument should include a program for testing the performance of the instrument against its specifications. If accuracy and precision over the dose or dose-rate range that will be used are the only important specifications, this will require one, or more, calibrated sources. If other specifications are also important, more complex testing will be necessary. Depending on the specifications, this could require a facility that includes a chamber to test the effects of pressure, temperature and humidity. It may also require a method for testing the detector over a large dynamic and a large energy range. Verifying the long-term stability and the lifetime of the power supply may not be possible as part of a n acceptance test, but should be warranted by the manufacturer. 10.2 Calibration

Calibration of instruments requires consideration of a number of factors that depend on the use of the instrument. These factors are listed here and explained below (NCRP, 199:Lb; Wagner, 1983): 1. 2. 3. 4. 5. 6.

level of calibration response of the instrument uncertainty in the calibration process frequency of calibration calibration facility design choice of sources

The calibration of survey instruments for the assessment of ionizing radiation fields and radioactive surface contamination was thoroughly discussed in NCRP Report No. 112 (NCRP, 1991b). Only a brief summary will be given here. The calibration requirements of other types of instrumentation that are used in a radiation safety program are mentioned briefly in this Section, but are not covered in detail. Calibration refers to the determination and adjustment of instrument response in a particular radiation field of known intensity. Three levels of calibration are full characterization, specific acceptance, and routine calibration. Full characterization is a process that would normally be done by the manufacturer of an instrument to determine the response of the instrument as completely as is possible. It includes evaluations of the energy dependence of the response, exposure linearity, the

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97

effects of radiation types other than those for which the instrument is designed, environmental influences, nonionizing radiations, instrument orientation, sensitivity to mechanical shock, dose-rate dependence, and angular dependence. Specific acceptance refers to the need to determine the instrument response, or to calibrate the instrument, under environmental or other conditions that are different from those that are encountered in normal operations. This may be necessary if the instrument is to be used under extremes of temperature, pressure, humidity, electric or magnetic fields. Routine calibration is the level that is appropriate for most individuals or institutions. This is applicable when the instrument is being used for measuring the radiation under the conditions for which it was designed and characterized. This level commonly involves the determination and adjustment of instrument response in known radiation fields from sources that cover the energy and dose-rate range in which the instrument will be used. The response of an instrument refers to the indication a n instrument gives on some device when it is exposed to a radiation source or field. This may be an electrical current or a voltage pulse, but the indication on the device is usually a n exposure rate, an air kerma rate, a dose rate, a dose equivalent rate, or a counting rate. Calibrations should be made in radiation fields that are specified in the units for which the particular instrument readout is designed. Typically the instrument response should be determined for various conditions of intensity, radiation energy, and orientation in the field and environment. Uncertainties in the calibration process will be both random and systematic and will involve uncertainties in the calibration of the radiation sources, or fields, as well as uncertainties in the instrument positioning in the field and the instrument response. The random uncertainties can be estimated and treated by standard statistical techniques. The systematic uncertainties should be kept as small as possible by careful procedures and elimination of biases. NCRP Report No. 112 (NCRP, 1991b) recommends that survey meters and surface contamination meters be calibrated to indicate the true dose-related quantity or radioactive contamination with overall uncertainties in the range from 10 to 35 percent, depending on the intended use of the instrument. Counting instruments that are used for analyzing environmental samples or instruments that are used for area, environmental or personal monitoring may have calibration uncertainties that fall in the same range (Budnitz et al., 1983; NCRP, 1985b; Wagner, 1983).

98 / lo. RADIATION SAFETY INSTRUMENTATION The frequency a t which instruments are calibrated will depend on the use, purpose and required accuracy of the instrument. Historical records of instrument performance may also be useful in deciding on calibration frequency. In almost all cases, an instrument should be calibrated a t least once per year, and the frequency should be higher for those instruments that are used most often. Each time that an instrument is repaired or fails a performance check, a calibration should be made before the instrument is returned to service. The calibration facility should be designed for the specific requirements of the instruments that are used and the sources that are measured a t a particular institution, unless the facility is established to provide calibration services to clients outside the institution. Care should be given to controlling the interfering background and scattered radiation. The sources should be carefully chosen to provide the appropriate ranges of dose rate and energy. Proper equipment should be available to assure calibration of the radiation fields and to provide adequate safety for the staff. The staff should be highly qualified and well trained in the operating characteristics of the instrumentation as well as in the techniques and procedures of instrument calibration. 10.3 Instrument Maintenance

Instrument maintenance requirements include establishing criteria for: 1. frequency of maintenance operations 2. performance checks 3. power supply checks

The frequency a t which maintenance is performed depends on the use of the instrument and the accuracy required. Usually it is sufficient to repair the instrument when it is damaged or when a component fails. However, in some cases, especially if the instrument is the only one available, a preventive maintenance inspection schedule should be established to limit equipment failures. A procedure should be established for checking the instrument's performance on a regular basis. Ideally this check should be made prior to each use of the instrument, or daily if the instrument is used frequently. This check can be made using a small sealed source or a counting standard. If the instrument fails the

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performance check, it should be sent for repair and recalibration immediately. A performance check will usually be sufficient to test that the power supply is functioning properly. However, the power supply for almost all survey instruments is a battery, and generally there is a built-in battery test function. This should be used to check the battery prior to using the instrument. Fixed area monitors and counting instruments may have high voltage power supplies that operate off the main power line. If a stable voltage supply is needed, the power supply line should include a surge protector and a voltage regulator. 10.4 Use of Instruments and

Acceptable Uncertainty Radiation measurement instrumentation can be characterized by its use, and the type of instrument determines its use. These considerations can be summarized as follows: 1. Use categories personal dosimetry radiation protection surveys area monitors stationary monitors environmental monitors 2. Instrument type pulse counters ionization chambers scintillators solid state detectors The choice of instrument and the acceptable uncertainty in its calibration depends on the purpose for which the instrument will be used. For personal monitoring performed by using either personal dosimeters or by integrating rahation survey measurements, NCRP has recommended accuracy criteria based on expected exposures: 1. a t levels near the annual limit, a measurement accuracy of +30 percent should be achieved 2. at levels below one-fourth of the annual limit, a lower level of accuracy (e.g.,a factor of two) is acceptable

100 / lo. RADIATION SAFETY INSTRUMENTATION 3. for levels of exposure higher than the annual limit, an accuracy of a t least 220 percent is acceptable (NCRP, 1978b) I n the same report, NCRP also recommended that the precision of such measurements be within 4 0 percent. ANSI has also recommended performance criteria for personal dosimeters that specify acceptable levels of accuracy (bias) and precision that depend on the dose level (ANSI, 1993). For instrumentation to be used for conducting surveys, NCRP Report No. 112 (NCRP, 1991b) specifies the accuracy acceptance criteria in photon, beta radiation, neutron fields, and for surface contamination measurements. These recommendations are summarized in Table 10.1. NCRP Report No. 112 also provides specific guidance on the methods to quantify uncertainties associated with survey measurements. ANSI also provides performance criteria for radiation protection survey instrumentation (ANSI, 1989b; 1989~). For instrumentation intended for use as area monitors, both the magnitude of acceptable levels of calibration accuracy and precision and their relationship to dose limits as specified in NCRP Report No. 57 (NCRP, 1978b) are appropriate. That is, when the equivalent dose rate for a work area approaches the level that would suggest that it be classified in the "moderate" dose category (1mSv < E < 50 mSv), the measurement should be about *30 percent. If the work area dose category classification is "minimum," ( E I1mSv) a measurement uncertainty of up to 100 percent can be tolerated. If the work area dose category classification is "high" ( E > 50 mSv), "very high" ( E > 250 mSv), or "extreme" (E > 1,000 mSv), the measurement uncertainty should be less than *20 percent [see Table 6.1 (NCRP, 1986)l.

10.5 Selection of Instruments for

Various Applications 'I'ypical instrumentation for radiation safety programs includes survey meters for area and contamination surveys, stationary area and airborne radionuclide monitors, detectors that are worn by individuals for monitoring personal dose, and detection systems that are used to analyze environmental air, water, soil and vegetation samples for radioactive materials. There are two categories of instruments:

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1. those that measure ambient radiation fields

2. those that measure the radioactivity of a sample or on a surface

A discussion of the instruments that are appropriate for use in radiation field measurements can be found in NCRP Report No. 57 (NCRP, 1978b). Instruments that are useful for environmental radiation measurements and sample analysis are described in TABLE10.1-Summaly of NCRP recommendations concerning radiation survey instrumentation. Instrument Application

Short-Term Stability

Accuracy Acceptance Criteria

Dose-related measurements in photon radiation fields

*5% at H > 1 mSv h-I *lo% at 0.05 < H > 1mSv h-l *20% at H < 0.05 mSv h-'

Dose-related measurements in neutron radiation fields

Same as above

*20%a +30%afor a rate <0.02 mSv h-'

Dose-related measurements in beta radiation fields

Same as above

Product of calibration factor and instrument reading must yield true dose rate with overall uncertainty within *20%a. For p energiesb< 0.03 MeV, uncertainty may be i30%

Assessment of surface contamination

*20%

*20%awhen corrections are provided with the instrument rt30%afor a rate <10 pGy h-I

a At 95 percent confidence level. b~eta-particleenergies determined at the position of the instrument rather than at the source.

102 / lo. RADIATION SAFETY INSTRUMENTATION NCRP Report No. 50 (NCRP, 1976b). Extensive information about various detector and instrumentation systems can be found in Attix (1972), Attix and Roesch (1966), Attix and Tochilin (19691, Kase et al. (1985; 1987; 1990) Knoll (1989), and Price (1964). Radiation safety surveys to measure exposures in photon radiation fields should be made using air ionization chamber instruments. When these instruments are calibrated in terms of the air kerma (exposure) or dose to air, their response is almost independent of the photon energy over the range that is normally encountered. Ionization chambers can also be designed to cover a very wide range of dose rate, and operate accurately in pulsed radiation fields. Care must be taken, however, to be certain that charged particle equilibrium is achieved at the higher energies, by providing sufficient wall thickness, and that excessive attenuation of photons does not occur at low energies, by providing a thin entrance window. These two conditions may not be satisfied with the same instrument. For measurements at normal environmental radiation levels, the chamber must be made large or must be pressurized to achieve adequate sensitivity. Organic or inorganic scintillation detectors can be used to measure photon radiation at very low levels. The response of organic scintillators is relatively independent of photon energy. Generally, inorganic scintillators have a response that is energy dependent, but if the energy spectrum of the radiation is known, even inorganic scintillators can be calibrated to measure dose. Scintillation detectors are useful in survey meters for measuring very low-energy radiation because of their high sensitivity to low-energy photons. Pulse counting instruments, such as Geiger-Miiller counters, are very sensitive and useful for surveys in photon radiation fields when an indication of the presence of a radiation field and its relative magnitude is needed. In general, an instrument of this type should not be used for dose measurements unless it is specifically calibrated for the photon energy spectrum that is to be measured. Care must be taken when measuring in pulsed radiation fields. Because the detector will discharge with each pulse, it cannot give an accurate indication of the radiation field. In fact, i t is possible for the detector to saturate and for the instrument to indicate zero when in an intense pulsed radiation field. Measurements of beta radiation fields can usually be made using the same instruments that were described for measuring photon radiation fields. The appropriateness of the different types of detectors is the same for beta radiation as for photon radiation.

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However, the results of the measurement will depend on the survey and measurement technique. Because of the short range of the beta particle, if dose or dose rate is needed, the position of the instrument in relation to the source must be the same (or nearly the same) as the conditions under which the instrument was calibrated. It is also necessary that the source be identified and that the instrument be calibrated in a nearly identical beta-ray spectrum [see NCRP Report No. 112 (NCRP, 1991b)l. Area surveys for neutrons in neutron or in mixed neutron and photon radiation fields require instruments that are especially designed to detect neutrons efficiently while being relatively insensitive to photons. These instruments usually have either a proportional counter or a scintillator as the detector. They respond to neutrons that have been thermalized by passing through a specially designed moderator that surrounds the detector. The moderator design will determine the energy dependence of the detector response. Thus, it is important to calibrate these instruments by using a neutron spectrum that is representative of the spectrum that is to be surveyed. Since these instruments are pulse counters, they can be saturated by intense photon pulses in a pulsed radiation field. Care must be taken to be certain that the reading obtained truly represents the dose rate in the radiation field and not simply the pulse rate of the source. Usually neutron survey instruments of the type described indicate the dose equivalent rate. This means that they have been designed to respond across the neutron energy spectrum in accordance with an accepted energy-dependent fluence-to-dose equivalent conversion. Since these relationships change with time as the relationship between dose and dose equivalent (or equivalent dose) is refined, it is, in principle, better to measure the neutron fluence and energy spectrum. The fluence measurement is relatively simple to do with the instruments that have been described, but the spectrum measurements are very difficult. Regardless of the difficulty, if neutron exposures are likely to contribute significantly to the dose to workers, measurements of the neutron spectra should be made. Techniques for making such measurements have been described by Cross and Ing (1987). S u ~ e yfor s radioactive contamination require instruments that are very sensitive to beta or alpha particles. Usually these will be instruments that have pulse counter detectors such as proportional counters, Geiger-Miiller counters or scintillators and indicate a count rate, rather than a dose rate. Instruments that are designed to detect alpha radiation require a very thin window so that the

104 / lo. RADLATION SAFETY INSTRUMENTATION alpha particle can penetrate to the sensitive volume. The readout electronics should incorporate a pulse-height discrimination circuit so that the instrument can be made insensitive to beta and photon radiation. Silver-activated ZnS scintillators are very insensitive to photons and are good alpha particle detectors. Geiger-Muller counters and organic scintillators are very good detectors for beta-particle contamination surveys. The GeigerMuller counters require a thin window, but the window can be made thick enough to exclude alpha particles. However, some very low-energy beta-particle emitters may not be detected a t all (e.g., 3~), or a t low enough levels (e.g., 14c), with these instruments, and other techniques may be required to detect contamination from such low-energy beta emitters. Instrumentation that is used for stationary monitoring of work areas, airborne radionuclides, effluent streams and the general environment will make use of the same detectors that are used in survey instrumentation. Stationary area monitors should use ionization detectors, scintillators or other solid state detectors operated in the current mode when measuring photon radiation fields to avoid the problems of saturation in pulsed or very high-dose rate fields. Proportional counters are appropriate for neutrons. The readouts should be continuously recorded so that both short- and long-term changes in the radiation field can be documented. There is seldom any need to have stationary area monitors specifically for alpha or beta radiation. Passive detectors can also be used for monitoring the general radiation environment in the work area as well as a t the boundary of, or in the vicinity of the facility. Typically the detectors that are appropriate for this type of monitoring will be small, resistant to changing environmental conditions and relatively independent of the energy of the radiation. Thermoluminescent detectors (TLD), such as LiF or CaS04, are very useful for monitoring ambient pho~ track ton radiation. Combinations of 'L~F and 7 ~ or ietchable detectors can be used for neutron radiation measurements (Griffith and Tommasino, 1990). Recently, new detectors have been developed that may also be useful for this purpose, e.g., alanine dosimeters that are interpreted by measuring their retained electron spin resonance signal (Hansen et al., 1987; Regdla and Deffner, 1982), and superheated liquid drop (bubble) detectors that are sensitive to neutron radiation (Apfel, 1979; 1992; Ing and Birnboim, 1984; Ipe et al., 1990; Schulze et al., 1992). Airborne radioactive particulates are usually monitored by passing air through a filter and measuring the activity on the filter.

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105

The design of the collection and measurement system depends on the radiation that is to be detected. If alpha particle-emitting radionuclides are to be detected, the particles should be collected in a thin layer on the surface of a membrane filter or impactor plate. A semiconductor surface barrier detector is an efficient detector that can discriminate against alpha particle-emitting background, such as radon decay products, using a pulse height analyzer. Material that is collected on a fiber filter paper can be monitored for airborne beta-particle emitters by using a Geiger-Muller or proportional counter. Certain radionuclides, such as 1 2 5 ~and 3 ~ require specialized monitoring techniques that are less amenable to continuous monitoring. Iodine may be trapped on activated carbon or other appropriate absorbent fdters. Generally the photon emission will be detected. Tritium can be monitored by passing air through a n ionization chamber, but the sensitivity is low and the reading must be corrected for background interference. Stationary monitoring methods can be used to obtain an indication of a release of airborne radionuclides into an environment, but they are usually not quantitative. Accurate quantification of airborne radionuclides requires that samples be taken that can be related to a precise volume of air, and radioactivity of the samples should be measured using a calibrated counting system. Continuous monitoring of liquid effluent streams can be done by immersing an appropriately designed Geiger-Muller or scintillation detector into the waste stream. Alpha particle emitters and low-energy beta-particle emitters will not be detected, and the overall sensitivity is low. Buildup of contamination on the detector will cause significant uncertainty in the result. Again, representative sampling of the waste stream and measuring the sample in the laboratory is the preferred method for detecting radionuclides in liquid effluent. A treatment of laboratory measurement of air, liquid or other environmental samples is beyond the scope of this Report. References relevant to this subject include NCRP Report No. 57 (NCRP, 197813) and NCRP Report No. 58 (NCRP, 198513). Instrumentation systems would include proportional counters, scintillators and semiconductor detectors that are connected to counting electronics or pulse height analyzers. Instrumentation that can be used for personal radiation monitoring includes ionization chambers, Geiger-Muller detectors, TLD, etchable track detectors, film and new materials such as alanine and superheated liquid drop (bubble) detectors. Ionization chambers, Geiger-Muller detectors and bubble detectors can be built into

,

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lo. RADIATION SAFETY INSTRUMENTATION

instruments that provide a direct and real-time indication of the radiation dose rate or the radiation dose that has been accumulated. TLD, track detectors, film, alanine and some bubble detectors accumulate a signal as the integrated dose increases and must be interpreted by using a technique that requires some processing of the detector, e.g., heating for TLD, developing for film, etching and counting tracks for track detectors, acquiring a spin resonance spectrum for alanine, and counting bubbles in the bubble detector. 10.6 Records for an Instrument Program

Detailed guidelines for records that should be maintained as part of a radiation safety instrument program are specified in NCRP Report No. 114 (NCRP, 1992). It is important that adequate records be kept to document the specification of instrumentation for each specific purpose, the calibration of the instrumentation, the maintenance and repair of the instrumentation, as well as an instrument inventory. For institutions or facilities that have large numbers of instruments, these records should be kept in a computer data base and linked in such a way that all the information concerning a particular instrument can be retrieved easily. Specification records should include the type of detector, energy range, dynamic range, sensitivity, accuracy and precision, response time, directional response, mixed radiation field response, and any restrictions on the operating environment of the instrument. Calibration records should include information about the calibration facility, such as a general description, standard instrumentation, radiation sources, calibration information pertaining to the sources, and the results of any studles of background or radiation scattering that are applicable to the facility. Instrument calibration records should include the instrument identification, the date of the calibration, the sources used, the individual who performed the calibration, the environmental conditions, and the results including any adjustments that were made. Maintenance records should include the instrument identification, the date of repair, a statement of the problem and corrective action, the individual who made the repair, and whether a recalibration was indicated. Information about instrument reliability is also useful to provide guidance for replacing instruments. Inventory records should include the instrument identification, the dates the instrument was placed into or removed from service, and the location or person to whom the instrument is assigned.

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Finally, there should be records of the radiation survey and monitoring program. These records should include the areas to be surveyed and the radiations expected, the areas in which stationary monitoring is required and the radiations expected, the areas that require contamination monitoring, airborne radionuclide or effluent monitoring and the radiations expected. The requirements of the personal dosimetry program should include the types of detectors used, the detector exchange frequency, the lower limit of detection, and the individuals who are to be monitored. Records of the results of the survey and monitoring program should also be maintained in accordance with the guidelines given in NCRP Report No. 114 (NCRP, 1992). 10.7 Recommended Additional Reading

"A bubble-damage polymer-detector for neutrons," Ing, H. and Birnboim, H.C., Nucl. Tracks. Radiat. Meas. 8,285-288, 1984. "Current and future instrument designs for monitoring photon radiations in terms of ambient and directional dose equivalent," Burgess, P.H., Rad. Prot. Dos. 12,211-214, 1985. "The development of a reference instmment for the direct determination of dose equivalent from beta radiation at various depths in tissue," Francis, T.M., Rad. Prot. Dos. 12,219-222, 1985. "Dosimeter Corporation's RAM Survey Instrumentation System," Bunker, A.S., Rad. Prot. Manag. 4, No. 6,41-49, 1987. Measurement of Dose Equivalents from External Photon and Electron Radiations, ICRU Report 47, International Commission on Radiation Units and Measurements, Bethesda, Maryland, 1992. "Precision of radiation monitoring measurements," Selvidge, J., Health Phys. 30,479-484,1976. Radiation Protection Instrumentation and Its Application, ICRU Report 20, International Commission on Radiation Units and Measurements, Bethesda, Maryland, 1971. T h e response of the RO-2 survey meter to mixed radiation fields," Shonka, J.J., Tschaeche, AN.,Tomblison, M.R., Gibeault, G.L., Moon, U.Y., McCoy, G.C. and Schrader, B.J., Rad. Prot. Manag. 6, NO.6,69-78, 1989.

11. Planning for Radiation Emergencies Every facility should have a n emergency plan, which may be very simple or very complex depending on the facility. Nevertheless, in developing an emergency plan there are a number of components that are common to all plans regardless of the complexity of the facility operation. These basic components are the evaluation of the types of accidents that might occur, the assignment of responsibilities for those who respond to accidents, the training of the responders and other persons, the detailed procedures that should be followed, the coordination with outside agencies, and provisions for periodically testing the plan. Emergency planning is discussed in detail in NCRP Report No. 111 (NCRP, 1991a) and the process can be classified as follows: 1. 2. 3. 4. 5.

development of the emergency plan preparation of implementing procedures classification of emergencies practical considerations evaluation and implementation

11.1 Development of the Emergency Plan

An emergency plan must be supported by the level of management of the facility who has the authority to commit resources and to implement policy to assure that there is an appropriate level of preparedness. Management should appoint an emergency coordinator to supervise the development and maintenance of the emergency plan. The emergency coordinator should, in addition, be the contact person for outside assistance organizations such as fire and police. The emergency plan should be useful for any type of emergency, not only those that involve radiation. A radiation emergency might also include fire, hazardous material releases and physical injuries that need to be addressed either separately or in conjunction with radiation.

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The organizational structure, the lines of authority and the functions of all individuals who will respond to a n emergency should be clearly defined. Duties of key individuals should be designated so that the required level of response can be determined quickly. Key individuals should be assigned to address all of the following items depending on the complexity of the plan: 1. radiation safety 2. security and traffic control

3. 4. 5. 6. 7. 8.

public information fire safety and hazardous substance control physical plant services medical services legal counsel

The key individuals responsible for these functions should report to the Emergency Director who exercises command and control in the event of a declared emergency. The Emergency Director may be the same person as the Emergency Coordinator who is responsible for developing the plan. 11.2 Preparation of Implementing Procedures

If the emergency plan is complex, it should be implemented through Emergency Plan Implementing Procedures (EPIP). These are documented instructions that describe the actions that are necessary to achieve the objectives of the plan. EPIPs should be written to cover the range of credible emergencies. They should cover the actions needed both during the emergency and during restoration following the emergency. Each EPIP should contain: 1. a statement of the purpose of the EPIP 2. identification of the person or group who is responsible for the EPIP 3. identification of the communication techniques to be used 4. a description of the action sequence required to achieve the purpose 5. a description of the prerequisites to the specified actions 6. the precautions necessary and the limitations of the prescribed tasks 7. guidelines to be followed in the exercise of judgment

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8. training requirements 9. copies or examples of the forms to be used 10. sign-off sheets, checklists and data sheets

In general, EPIPs should be prepared for the emergency director, security and police personnel, RSO,public information officer, plant services director, institutional fire marshal, medical officer, and legal counsel. It is important that each EPIP describe the initial training and periodic retraining programs that are required for primary and backup persons who will be responsible for the EPIP. It is the responsibility of the site management to make available to the emergency response teams the facilities and equipment that are required by the EPIPs. Equipment that may be needed includes radiation detection instrumentation, sampling devices, personal dosimeters, personal protective equipment, decontamination supplies and communications systems. These should all be ready and operable. Measurement and sampling equipment should be calibrated and consistent with the requirements of the plan. 11.3 Classification of Emergencies

To assist in planning for the consequences of an emergency, it is appropriate to categorize and to identify the location of sources of radiation. These may include sealed or encapsulated sources, unsealed sources and radiation-producing machines. Classifying the source aids in evaluating the type of radiation hazard that may exist in the event of an emergency. The following are examples of possible hazards: 1. external whole or partial body exposures 2. exposures from skin contamination

3. internal exposures from ingestion or inhalation 4. internal exposures from puncture or laceration 5. internal and/or external exposures from submersion in a radioactive medium 6. environmental release or contamination A large variety of potential emergency situations can be postulated, but it is important that plans be limited to credible events to ensure appropriate levels of response. As an illustration, if skin contamination is not credible, the plan should not include a hazard classification for skin contamination. If, at a later date, the facility

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acquires a source that could result in skin contamination in a n emergency, the hazard classification scheme should be revised to include it. Associated hazards such as infectious agents and toxic materials should be included in the classification scheme when appropriate. Decontamination agents that may be desirable for radiation safety purposes but undesirable for infectious agents should be evaluated. The seriousness of a n emergency should also be classified to aid the Emergency Director and the emergency team in implementing the appropriate parts of the plan. The following questions should be asked when classifying the seriousness of an emergency: 1. Will the emergency be confined to a single location? 2. Will parts of the facility be affected that normally would be free of radiation or radioactive material? 3. Are radiation exposures to affected individuals or the emergency team likely to exceed applicable effective dose guidelines? 4. Will the emergency involve persons not associated with the facility such as community fire fighters? 5. Is it likely that there will be radiation exposures to the general public? 6. Is extensive contamination or the release of a large quantity of radioactive material likely?

11.4 Practical Considerations

Some practical considerations are important when developing and implementing an emergency preparedness plan. Among the first tasks is the assessment of the magnitude of the event. The h i tial assessment may be made by inexperienced individuals. Additional information may be needed before a decision can be reached to declare a n emergency and to decide what parts of the plan should be activated. A well-informed judgment will be necessary if the event does not fall into the classification scheme. In all cases, there must be timely communication of site conditions and evaluation team findings to the decision makers. Following the emergency, restoration and recovery will depend on the events that took place. Access control may be necessary and dose assessments may be required. Cleanup activities may not be a part of the plan, but as a practical matter, need to be contemplated. In addition, there should be some practical mechanism for

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declaring that the emergency is over and for putting the facility back under normal control.

11.5 Evaluation of the Plan Once the emergency plan has been drafted, it should be reviewed, evaluated and tested. If there is a master emergency plan a t the facility, the radiation emergency plan should be incorporated into, and function within, that plan. To be effective, the completed plan must be explained to leaders of all functional groups. Their advice and input should be solicited on the section of the plan in which they will participate. The draft plan should be approved by top-level management and by the external agencies that will participate in it. When i t is complete, the plan should be thoroughly tested. It may be desirable to test portions of the plan independently to be certain that the functional groups are working properly. A full scale exercise should be designed to simulate a plausible emergency. During the exercise selected evaluators should observe and comment on the test. Participants should be asked to comment on the strengths and weaknesses of the plan. Following the exercise, comments from the evaluators and participants should be reviewed and discussed to determine if there is a need to modify or amend the plan. This should be done by means of a critique session immediately after the exercise. Problems with the plan should be discussed by all participants. The emergency plan should be tested by using both announced and unannounced exercises. In most cases the notification sequence and communications sections can be tested without an announcement. This is particularly useful when telephone numbers are likely to have changed. Larger exercises require planning involving the participants to ensure that the objectives are achieved. The emergency scenario that is used to test the plan should:

1. 2. 3. 4. 5. 6.

be specific provide a true test of the plan and organization include activities that will facilitate the evaluation incorporate off-site participants as needed address regulatory requirements list the actions of participants that will be demonstrated

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7. be within the scope of the potential emergency classification scheme 8. test the communication network 9. test the level of training of the emergency responders Following the exercise and the critique session, and after a review of the strengths and weaknesses of the plan, the Emergency Coordinator should correct any basic problems and modify the plan as necessary. Any identified weaknesses should be monitored until a subsequent exercise demonstrates that corrective actions have been effective.

Glossary activation: The process of inducing radioactivity. administrative dose guidelines: The predetermined value of radiation dose to workers, below the dose limit, which triggers a specific course of action when the value is exceeded, or is expected to be exceeded. ALARA (aslow as reasonably achievable):A principle of radiation protection philosophy that requires that exposures to ionizing radiation should be kept as low as reasonably achievable, economic and social factors being taken into account. The ALARA principle is equivalent to the principle of optimization defined by the ICRP which states that protection from radiation exposure is optimum when the expenditure of further resources would be unwarranted by the reduction in exposure that would be achieved. alpha particles: A charged particle having a mass and charge equal to a helium nucleus (i.e., two protons and two neutrons) emitted from the nucleus of a n atom. annual reference level of intake (ARLI):The activity of a radionuclide taken into the body during a year that would result in a committed effective dose of 20 mSv to an individual represented by Reference Man. The ARLI is expressed in becquerels (Bq). area monitor: A radiation detector designed to measure the radiation levels in a specified location. beta particle: A negatively charged particle indistinguishable from a n electron that is emitted from the nucleus of a n atom as a consequence of radioactive decay. bioassay: A technique used to identify, quantify andlor specify the location of radionuclides in the body by direct (in vivo) or indirect (in vitro) analysis of tissues or excretions from the body. collective effective dose: The sum of the individual effective doses to a specified group or population. contamination: Radioactive material suspended in air or deposited in any area or on any surface where its presence is unwanted or unexpected.

committed effective dose: The time integral of the effective dose rate from an intake of a radionuclide. Unless specified otherwise, the time interval is taken to be 50 y for those exposed in the workplace and 70 y for members of the public. criticality: The point at which a nuclear fission reaction becomes capable of sustaining a chain reaction. decommission: Close a facility and prepare its buildings and land for release to unrestricted use. derived reference air concentration (DRAC): The ARLI of a radionuclide divided by the volume of air inhaled by Reference Man in a working year (i.e., 2.4 x lo3 m3). The unit of the DRAC is Bq m-3. detriment: The total harm that is expected to be experienced by an exposed group and its descendants as a result of the group's exposure to radiation. The expected harm includes fatal and nonfatal cancer, severe hereditary effects, and life shortening. dose (Dl: The quotient of dE by dnz,where d& is the mean energy imparted by ionizing radiation to matter of mass dm, thus the dose, D is:

The unit of dose is the gray (Gy); formerly the unit was the rad (1 Gy = 100 rad). The term "dosen is often used in an informal way to mean effective dose, equivalent dose, or committed dose depending on the context. dosimeter: A device, instrument or system usually worn by an individual to determine his or her personal dose equivalent. effective dose: The sum over specified tissues of the products of the equivalent dose in a tissue (T)and the weighting factor for That is, that tissue (wT).

where E is the effective dose and HT is the equivalent dose to tissue T. The unit of effective dose is the sievert (Sv); formerly the unit was the rem (1Sv = 100 rem). effluent monitoring: The measurement of radioactivity in air, liquids and solid material leaving a facility. emergency: A sudden, urgent, usually unforeseen occurrence or occasion requiring immediate action.

116 / GLOSSARY engineered controls: A system of design features that are intended to prevent or limit exposures to radiation or releases of radioactive materials. environmental monitoring: The measurement of external dose or the amount of radioactivity in air, soil, water, plant and m i mal matter in areas outside the control or boundary of a facility. equivalent dose: A quantity obtained by multiplying the average absorbed dose in a tissue or organ by a radiation weighting factor (wR)to allow for the different effectiveness of the various ionizing radiations in causing harm to tissue. The unit of equivalent dose is the sievert (Sv); formerly the unit was the rem (1Sv = 100 rem). external dose: The dose received from a source of radiation outside of the body. fail safe: A system design that ensures that any system failure will always leave the facility in a safe configuration. fissile: A descriptor of a nuclide that is capable of undergoing nuclear fission, usually following the absorption of a slow neutron. formal procedure: A process or task that is executed in accordance with written documentation that has been approved by competent authority and for which records of its changes and implementation are maintained. Geiger-Miiller counter: A gas-filled radiation detector most oRen used to detect the presence of low dose rate beta particles, x rays, or gamma rays. The detector is not appropriate for use with pulsed radiation sources or when the type or energy of the radiation is to be determined. gamma rays: Short wavelength electromagnetic radiation, usually with a n energy greater than 100 keV, emitted from a n atomic nucleus as a result of radioactive decay. gloved box: A sealed container with controlled access and ventilation, usually having a viewing window and ports covered by gloves to provide access for work inside the box. grab sample: A sample of limited volume taken a t random or a t preselected frequencies. hot cell: A heavily shielded enclosure for handling, processing or storing highly radioactive sources. internal dose: The dose received from a source of radiation inside of the body.

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isokinetic sampling: Sampling a n effluent without perturbing the dynamic flow of either the effluent or particulates within the effluent. justification: The part of the decision-making process in which the options that are expected to do more good than harm are identified. limit: In radiation protection, the level of dose established by authoritative or consensus bodies above which the consequences to an individual would be regarded as unacceptable. monitoring: Continuous or periodic determination of the amount of radiation or radioactivity present in a specified area or volume of effluent. negligible individual dose: A level of effective dose to an individual per source or practice that may be ignored. This term was defined in NCRP Report No. 127 and its recommended value is 0.01 mSv. objective health detriment: All stochastic health effects for which quantitative estimates of the probability of occurrence as a function of radiation dose have been derived from actual exposed populations. occupancy factor: The factor used in dose projection calculations for shielding design or other purposes to account for the fraction of time that a space will be occupied by any single individual. occupational exposure: Exposures to individuals that are incurred in the workplace as the result of situations which can reasonably be regarded as being the responsibility of management (exposures associated with medical diagnosis or treatment are excluded). personal dose equivalent [H,(d)l: The equivalent dose determined a t an appropriate depth, d, in the body. This depth is usually taken to be 10 mm for penetrating radiation and 0.07 mm if the individual is expected to be exposed to nonpenetrating radiation. Hp(lO) and Hp(0.07) are also referred to as the "deep dosen and "shallow dose" respectively. The unit for the personal dose equivalent is the sievert (Sv); formerly the unit was the rem (1Sv = 100 rem). planned special exposure: A radiation dose that is authorized for a worker that will exceed occupational dose limits. The planned special exposure is considered only in exceptional circumstances when it can be justified and when alternatives that might prevent a worker from exceeding limits are unavailable or impractical.

118 1 GLOSSARY public: All persons who are not already considered occupationally exposed by a source or practice under consideration. qualified expert: In radiation protection, a person having the knowledge and training to provide advice regarding radiation protection principles, standards and measurements. Persons having relevant certification from the American Board of Health Physics, the American Board of Medical Physics, the American Board of Radiology, or the American Board of Industrial Hygiene may be considered qualified experts. radiation safety: Control of the sources of radiation and the exposure to radiation to protect people and the environment from unnecessary exposure and the deleterious effects of exposure to radiation. radiation work permit (RWP):With regard to radiation safety, a n authorization to perform a specific procedure that will involve the exposure of persons to radiation or uncontained radioactive material in a specified area of a facility. radionuclide: A radioactive species of atom characterized by its mass number, atomic number and sometimes its nuclear energy state (provided that the mean lifetime of that state is sufficiently long for the species to be observed). reference levels: See "administrative dose guidelines". Reference Man: A person with the anatomical and physiological characteristics defined in the report of the ICRP Task Group on Reference Man (ICRP, 1975). respiratory protection device or respirator: A device worn on the face or head to prevent the inhalation of toxic or radioactive materials. surveys: Periodic determination of the amount of radiation or the amount of radioactive material present in an area.

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IPE, N.E., DONAHUE, R.J. and BUSICK, D.D. (1990). "The active personnel dosemeter-Apfel Enterprises Superheated Drop Detector," Rad. Prot. Dos. 34, 157-160. JAEGER, R.G., Ed. (1975). Engineering Compendium on Radiation Shielding: Volume 11, Shielding Materials (Springer-Verlag, New York). KASE, K.R., BJARNGARD, B.E. and ATTM, F.H., Eds. (1985). The Dosimetry of Ionizing Radiation, Volume Z (Academic Press, New York). KASE, K.R, BJMNGARD, B.E. and ATTIX, EH., Eds. (1987). The Dosimetry of Ionizing Radiation, Volume ZZ (Academic Press, New York). KASE, K.R., B J r n G A R D , B.E. and ATTM, EH., Eds. (1990). The Dosimetry of Ionizing Radiation, Volume I I Z (Academic Press, New York). KATKREN, R.L.,YODER, R.L., DEBROSIERS, A.E., NISICK, N.P., HOWELL, W.P. and WAITE, OSCARSON, E.E., MULHEARN, O.R., D.A. (1980). "Radiological design," pages 5.3 to 5.25 in A Guide to Reducing Radiation Exposure to As Low As Reasonably Achievable (ALARA), DOE/EV/1830-T5 (National Technical Information Service, Springfield, Virginia). KNIEF, RA. (1993). Nuclear Criticality Safety: Theory and Practice (American Nuclear Society, LaGrange Park, Illinois). KNOLL, G.E (1989). Radiation Detection and Measurement, 2nd ed. (Wiley, New York). LEE, R. (1994).T h e effects of respiratory protection on worker e S ciency," Rad. Prot. Manage. 11,No. 4 , 2 8 4 2 . LESSARD, E.T., YrHUA, X., SKRABLE, K-W., CHABOT, G.E., FRENCH, C.S., LABONE, T.R., JOHNSON, J.R., FISHER, D.R., BELANGER, R. and LIPSZTEIN, J.L. (1987). Interpretation of Bioassay Measurements. Regulatory Commission, NUREG/CR-4884, 1J.S. Nuclear BNL-NUREG-52063 (National Technical Information Service, Springfield, Virginian). MCDERMOTT, H.J. (1985). Handbook of Ventilation for Contaminant Control, 2nd ed. (Butterworth Heinemann, Stoneham, Massachusetts). MCGINLEY, P.H. (1993). Design of Therapy Facilities in Hospital Health Physics, Eichholz, G. and Shonka, J., Eds. (Health Physics Society, McLean, Virginia). NCRP (1971). National Council on Radiation Protection and Measurements. Protection Against Neutron Radiation, NCRP Report NO. 38 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1976a). National Council on Radiation Protection and Measurements. Structural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV NCRP Report

124 /

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No. 49 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1976b). National Council on Radiation Protection and Measurements. Environmental Radiation Measurements, NCRP Report No. 50 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1977) National Council on Radiation Protection and Measurements. Radiation Protection Design Guidelines for 0.1-100 MeV Particle Accelerator Facilities, NCRP Report No. 5 1 (out of print). NCRP (1978a). National Council on Radiation Protection and Measurements. Operational Radiation Safety Program, NCRP Report No. 59 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1978b). National Council on Radiation Protection and Measurements. Instrumentation and Monitoring Methods for Radiation Protection, NCRP Report No. 57 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1980). National Council on Radiation Protection and Measurements. Management of Persons Accidentally Contaminated with Radionuclides, NCRP Report No. 65 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1983a). National Council on Radiation Protection and Measurements. Operational Radiation Safety-Training, NCRP Report No. 71 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1983b). National Council on Radiation Protection and Measurements. Radiation Protection and Measurements for Low Voltage Neutron Generators, NCRP Report No. 72 (National Council on Radiation Protection a n d Measurements, Bethesda, Maryland). NCRP (1985a). National Council on Radiation Protection and Measurements. General Concepts for the Dosimetry of Internally Deposited Radionuclides, NCRP Report No. 84 (National Council on Radiation Protection a n d Measurements, Bethesda, Maryland). NCRP (1985b). National Council on Radiation Protection and Measurements. A Handbook of Radioactivity Measurements Procedures, NCRP Report No. 58, 2nd ed. (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1986). National Council on Radiation Protection and Measurements. Radiation Alarms and Access Control Systems, NCRP Report No. 88 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1987). National Council on Radiation Protection and Measurements. Use of Bioassay Procedures for Assessment of Internal Radionuclide Deposition, NCRP Report No. 87 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1989a). National Council on Radiation Protection and Measurements. Radiation Protection for Medical and Allied Health Personnel,

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NCRP Report No. 105 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (198913). National Council on Radiation Protection and Measurements. Exposure of the U S . Population from Occupational Radiation, NCRP Report No. 101 (out of print). NCRP (1989~).National Council on Radiation Protection and Measurements. Medical X-Ray, Electron Beam and Gamma-Ray Protection for Energies Up to 50 MeV (Equipment Design, Performance and Use), NCRP Report No. 102 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1989d). National Council on Radiation Protection and Measuements. Screening Techniques for Determining Compliance with Environmental Standards, NCRP Commentary No. 3, rev. (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1990). National Council on Radiation Protection and Measurements. Implementation of the Principle of As Low As Reasonably Achievable (ALARA) for Medical and Dental Personnel, NCRP Report No. 107 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1991a). National Council on Radiation Protection and Measurements. Developing Radiation Emergency Plans for Academic, Medical or Industrial Facilities, NCRP Report No. 111 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (199113). National Council on Radiation Protection and Measuements. Calibration of Survey Instruments Used i n Radiation Protection fir the Assessment of Ionizing Radiation Fields and Radioactive Surface Contamination, NCRP Report No. 112 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1992). National Council on Radiation Protection and Measurements. Maintaining Radiation Protection Records, NCRP Report No. 114 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1993a). National Council on Radiation Protection and Measurements. Radiation Protection i n the Mineral Extraction Industry, NCRP Report No. 118 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (199313). National Cotmcil on Radiation Protection and Measurements. Limitation of Exposure to Ionizing Radiation, NCRP Report No. 116 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1993~).National Council on Radiation Protection and Measurements. Uncertainty in NCRP Screening Models Relating to Atrnospheric Transport, Deposition and Uptake by Humans, NCRP Commentary No. 8 (National Council on Radiation Protection and Measurements, Bethesda, Maryland).

126 / REFERENCES NCRP (1994). National Council on Radiation Protection and Measuements. Dose Control at Nuclear Power Plants, NCRP Report No. 120 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1995a). National Council on Radiation Protection and Measurements. Use of Personal Monitors to Estimate Effective Dose Equivalent and Effective Dose to Workers for External Exposure to Low-LET Radiation, NCRP Report No. 122 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1995b). National Council on Radiation Protection and Measurements. Principles and Application of Collective Dose i n Radiation Protection, NCRP Report No. 121 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1996). National Council on Radiation Protection and Measurements. Screening Models for Releases of Radionuclides to the Atmosphere, Surface Water, and Ground, NCRP Report 123 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NRC (1994). U.S. Nuclear Regulatory Commission. Guide to Hazardous, Radioactive, and Mixed Waste Generators on the Elements of a Waste Nuclear ReguMinimization Program, Information Notice 94-23 (U.S. latory Commission, Washington). O'DELL, D.R., Ed. (1974). Nuclear Criticality Safety, TID-26286 (National Technical Information Service, Springfield, Virginia). PAXTON, H.C. (1989). Glossary of Nuclear Criticality Terms, LA-11627-MS, UC714 (Los Alamos National Laboratory, Los Alamos, New Mexico). PAXTON, H.C., THOMAS, J.T., CALLIHAN, A.D. and JOHNSON, E.B. (1964). Critical Dimensions of Systems Containing U-235, Pu-239 and U-233, TID-7028 (Los Alamos National Laboratory, Los Alarnos, New Mexico). PLOG, B.A., Ed. (1988). 'Industrial ventilation," pages 475 to 503 in Funclamentals of Industrial Hygiene, 3rd ed. (National Safety Council, Chicago, Illinois). PRICE, W.J. (1964). Nuclear Radiation Detection, 2nd ed. (McGraw-Hill, New York). PRICE, B.T., HORTON, C.C. and SPINNEY, K.T. (1957). International Series of Monographs on Nuclear Energy. Volume 2, Radiation Shielding, TK 9210 P1 (Pergamon Press, Elmsford, New York). PROFIO, A.E. (1970). Radiation Shielding and Dosimetry (Wiley, New York). REGULLA, D.F. and DEFFNER, U. (1982). "Dosimetry by ESR epectroscopy of alanine," Int. J. Appl. Radiat. Isot. 33, 1101-1114. ROCKWELL, T., 111, Ed. (1956). Reactor Shielding Design Manual (D. Van Nostrand, Princeton, New Jersey).

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SCHULZE, J., ROSENSTOCK, W. and KRONHOLZ, H.L. (1992). "Measurements of fast neutrons by bubble detectors," Rad. Prot. Dos. 44, 351-354. SHONKA, J.J., SCHRADER, B.J., TSCHAECHE, A.N., GIBEAULT, G.L., CLARKE, G.W. and TOMBLISON, M.R. (1990). "Field estimation of beta protection factors," Rad. Prot. Manage. 7, 51-56. SHULTIS, J.K. and FAW, R.E. (1996). Radiation Shielding (Prentice Hall PTR, Upper Saddle River, New Jersey). SKRABLE, K.W.,SUN, L.C., CHABOT, G.E., FRENCH, C.S. and LA BONE, T.R. (1987). "Pseudo uptake retention functions for the whole body for estimating intakes from excretion bioassay data," Rad Prot Dosim. 18, 133-139. TOOHEY, R., PALMER, E., ANDERSON, L., BERGER, C., COHEN, N., EISELE, G., WACHHOLZ, B. and BURR, W., JR.(1991). "Current status of whole-body counting as a means to detect and quantify previous exposures to radioactive materials. Whole-Body Counting Working Group," Health Phys. 60, 7-42. UL (1990). Underwriters Laboratories. High Effiiency, Particulate, Air Filter Units, -586 (Underwriters Laboratories, Northbrook, Illinois). UNREIN, P. J., PELLETIER, C. A., CLINE, J. E. and VOILLEQU~,P. G. (1985). "Transmission of radioiodine through sampling lines," pages 116 to 125 in Proceedings of the 18th DOE Nuclear Airborne Waste Management and Air Cleaning Conference, Report Number CONF-840806 (U.S. Department of Energy, Washington). WAGNER, S.R. (1983). "Measurement and calibration in radiation protection," Rad. Prot. Dos. 4, 67-70.

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 infor-

mation 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 cany 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 h l l membership of the Council for careful review and approval before being published. The following comprise the current officers and membership of the Council:

THE NCRP /

President Vice President Secretary and Assistant Tkeasurer Assistant Secretary ~asurer

WILLIAMM. BECKNER CARL D. HOBELMAN JAMES F. BERG

Members S. JAMES ADELSTEIN LARRY E. ANDERSON LYNNR.ANSPAUGH JOHN W. BAUM HAROLD L. BECK MICHAELk BENDER B. GORDONBLAYLOCK BRUCEB. BOECKER JOHN D. BOICE,JR BOUVlLLE LESLIE A. BRABY JOHN W. BRAND DAVID BRENNER ANTONE L. BROOKS PATRICIAA. BUFFLER CHUNG-KWANG CHOU JAMES E. CLEAVER J. DONALDCOSSAIRT ALLENG. CROFF PAULM. DELUCA GAILDE PLANQUE SAFMS. DONALDSON WILLIAMP. DORNSIFE KEITHF. ECKERMAN MARC EDWARDS H. KEITHFLORIG T H O W F. GESELL ETHELS. GILBERT JOHN D. GRAHAM JOEL E. GRAY

RAYMOND A.

ERICJ. HALL NAOMIH. HARLEY WILLIAMR. HENDEE DAVID G. HOEL F.OWENHOFFMAN R.HOWE GEOFFREY DONALDG. JACOBS KENNETH R. KASE

AMY KRONENBERG

CHARLESE. LAND RICHARD LEGGETT HOWARDLIBER JOHN B. L1l"rLE RICHARD A. LUBEN ROGER O. MCCLELLAN BARBARA J. MCNEU CHARLESB. MEINHOLD FREDA. METTLE&J R CHARLESW. MILLER KENNETHL. MILLER DAVID S. MYERS RONALDC. PETERSEN JOHN W. POM'ON, SR ANDREW K. POZNANSKI R. JULIAN PRESTON G E N S. ROESSLER ~ MARVIN ROSENSTELN LAWRENCE N.ROTHENBERG HENRYD. ROYAL M~CHAEL T. RYAN JONATHAN M. SAMET STEPHEN M. SELTZER ROYE. SHORE KENNETHW. SKRABLE DAVIDH. SLINEY PAULSLOVIC LOUISE C. STRONG A. TELL RICHARD THOMAS S. TENFORDE JOHN E. TILL LAWRENCE W. TOWNSEND ROBERT L. ULLRICH R~CHARDJ. VETTER DAVIDA. WEBER F. WARDWHICKER CHRISG. WHLPPLE J. FRANK WILSON SUSAN D. WILTSHIRE MARVIN C. ZISKIN

129

130 / THE N C R P Honorary Members LAURISTON S. TAYLOR,Honorary President WARRENK. SINCWR, President Emeritus W. ROGERNEY, Executive Director Emeritus THOMASS. ELY RICHARDF.FOSTER HYMERL. FRIEDELL R.J. MICHAELFRY ROBERT0.GORSON A R w. GUY ~ JOHN W. HEALY BERNDKAHN WILFRIDB. MANN DADEW. MOELLER A. ALAN MOGHISSI KARL 2. MORGAN ROBERTJ. NELSEN

WESLEYL. NYBORG CHESTERR. RICHMOND HARALD H. ROSS] WILLIAWL. RUSSELL JOHN H. RUST EUGENEL. SAENGER WILLIAMJ. SCHULL J. NEWELLSTANNARD JOHN B. STORER ARTHUR C. UPTON

GEORGEL. VOELZ EDWARLIW. WEBSTER HAROLD0.WCKOFP

Lauriston S. lbylor Lecturers The Squares of the Natural Numbers in Radiation Protection Why be Quantitative about Radiation Risk Estimates? Radiation Protection-Concepts and Trade Offs From "Quantity of Radiation" and 'Dose" to "Exposuren and "Absorbed Dosen-An Historical Review How Well Can We Assess Genetic Risk? Not Very Ethics, Trade-offs and Medical Radiation The Human Environment-Past, Present and Future Limitation and Assessment in Radiation Protection lkuth (and Beauty) in Radiation Measurement Biological Effects of Non-ionizing Radiations: Cellular Properties and Interactions How to be Quantitative about Radiation Risk Estimates How Safe is Safe Enough? Radiobiology and Radiation Protection: The Past Century and Prospects fir the Future Radiation Protection and the Internal Emitter SWQ When is a Dose Not a Dose? Dose and Risk in Diagnostic Radiology: How Big? How Little? WARRENK. SINCLAIR Science, Radiation Protection and the NCRP R.J. MICHAELFRY Mice, Myths and Men ALBRECHTKELLERER Certainty and Uncertainty in Radiation Protection SEYMOUR ABRAHAMSON 70 Years of Radiation Genetics: Fruit Flies, Mice and Humans Radionuclides in the Body: Meeting the Challenge! From Chimney Sweeps to Astronauts: Cancer Risks in the Workplace

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135

The Council's activities are made possible by the voluntary mntribution of time and effort by its members and participants and the generous support of the following organizations: 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 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 of Radiologic Technologists American Society for Therapeutic Radiology and Oncology American Veterinary Medical Association American Veterinary Radiology Society 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 of Pennsylvania Consumers Power Company Council on Radionuclides and Radiopharmaceuticals Defense Nuclear Agency Eastman Kodak Company Edison Electric Institute Edward Mallinckrodt, Jr. Foundation

136 / THE NCRP EG&G Idaho, Inc. Electric Power Research Institute Federal Emergency Management Agency Florida Institute of Phosphate Research Florida Power Corporation Fuji Medical Systems, U.S.A., Inc. Genetics Society of America Health Effects Research Foundation (Japan) Health Physics Society 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 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 United States Department of Energy United States Department of Labor United States Environmental Protection Agency United States Navy United States Nuclear Regulatory Commission Victoreen, Inc. Initial funds for publication of NCRP reports were provided by a grant from the James Picker Foundation. The NCRP seeks to promulgate information and recommen-dations 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 from those interested in its work.

THE NCRP

/ 131

Currently, the following committees are actively engaged in formulating recommendations: Basic Criteria, Epidemiology, Radiobiology and Risk SC 1-4Extrapolation of Risk from Non-Human Experimental Systems to Man SC 1-6Basis for the Linearity Assumption SC 1-7Information Needed to Make Radiation Protection Recommendations for Travel Beyond Low-Earth Orbit SC 1-8Risk to Thyroid from Ionizing Radiation Structural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV Operational Radiation Safety SC 46-8Radiation Protection Design Guidelines for Particle Accelerator Facilities SC 46-10Assessment of Occupational Doses from Internal Emitters SC 46-11Radiation Protection During Special Medical Procedures SC 46-13Design of Facilities for Medical Radiation Therapy SC 57-10Liver Cancer Risk SC 57-14Placental Transfer SC 57-15Uranium SC 57-16Uncertainties in the Application of Metabolic Models SC 57-17Radionuclide Dosimetry Models for Wounds Radionuclides in the Environment SC 64-17Uncertainty in Environmental Transport in the Absence of Site Specific Data SC 64-18Ecologic and Human Risks from Space Applications of Plutonium SC 64-19Historical Dose Evaluation SC 64-20Contaminated Soil SC 64-22Design of Effective Effluent and Environmental Monitoring Programs SC 64-23Cesium in the Environment Biological Effects and Exposure Criteria for Ultrasound Radiation Protection in Mammography Guidance on Radiation Received in Space Activities Risk of Lung Cancer from Radon Hot Particles in the Eye, Ear or Lung Radioactive and Mixed Waste SC 87-1Waste Avoidance and Volume Reduction SC 87-2Waste Classification Based on Risk SC 87-3Performance Assessment SC 87-4Management of Waste Metals Containing Radioactivity Fluence as the Basis for a Radiation Protection System for Astronauts

132 / THENCRP SC 89

SC 91

SC 92 SC 93

Nonionizing Electromagnetic Fields SC 89-1 Biological Effects of Magnetic Fields SC 89-3 Extremely Low-Frequency Electric and Magnetic Fields SC 89-4 Modulated Radiofrequency Fields SC 89-5 Biological Effects and Exposure Criteria for Radiofiequency Fields Radiation Protection in Medicine SC 91-1 Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides SC 91-2 Dentistry Public Policy and Risk Communication Radiation Measurement

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 the 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 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 the NCRP maintains liaison are as follows: American Academy of Dermatology American Academy of Environmental Engineers American Academy of Health Physics American Association of Physicists in Medicine American College of Medical Physics American College of Nuclear Physicians American College of Occupational and Environmental Medicine American College of Radiology American Dental Association American Industrial Hygiene Association American Institute of Ultrasound in Medicine American Insurance Services Group American Medical Association

THE NCRP / 133

American Nuclear Society American Pharmaceutical Association American Podiatric Medical Association American Public Health Association American Radium Society American Roentgen Ray Society American Society of Health-System Pharmacists American Society of Radiologic Technologists American Society for Therapeutic Radiology and Oncology 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 Radiophamaceuticals Defense Special Weapons Agency Electric Power Research Institute Electromagnetic Energy Association 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 National Aeronautics and Space Administration National Association of Environmental Professionals National Electrical Manufacturers Association National Institute of Standards and Technology Nuclear Energy Institute Office of Science and Technology Policy Oil, Chemical and Atomic Workers Union Radiation Research Society Radiological Society of North America Society of Nuclear Medicine Society for Risk Analysis United States Air Force United States Army United States Coast Guard United States Department of Energy United States Department of Housing and Urban Development United States Department of Labor United States Department of Transportation United States Environmental Protection Agency United States Navy United States Nuclear Regulatory Commission United States Public Health Services Utility Workers Union of America

134 / THE NCRP The 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 the 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 the 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 a n opportunity to make suggestions on new studies and related matters. The following organizations participate in the Special Liaison Program: Australian Radiation Laboratory Central Laboratory for Radiological Protection (Poland) European Commission Federal Office for Radiation Protection (Germany) Health Council of the Netherlands Institute de Protection et de Surete Nucleaire (France) International Commission on Non-Ionizing Radiation Protection Japan Radiation Council Korea Institute of Nuclear Safety National Radiological Protection Board (United Kingdom) National Research Council (Canada) Russian Scientific Commission on Radiation Protection South &can Forum for Radiation Protection Ultrasonics Institute (Australia) The 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 Health Physics Services Amersham Corporation Commonwealth Edison Consolidated Edison Duke Energy Corporation Landauer, Inc. New York Power Authority Nuclear Energy Institute Southern California Edison Westinghouse Electric Corporation

NCRP Publications Information on NCRP publications may be obtained from the NCRP website (http://www.ncrp.com) or by telephone (800-229-2652) and fax (301-907-8768). The address is: NCRP Publications 7910 Woodmont Avenue Suite 800 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. 8

22

25 27 30 32 35

36 37 38

41 42

Title Control and Removal of Radioactive Contamination in Laboratories (1951) Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Rudionuclides in Air and in Water for Occupational Exposure (1959)Dncludes Addendum 1 issued in August 19631 Measurement ofAbsorbed 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 TherapeuticAmounts of Radionuclides (1970) Protection Against Neutron Radiation (1971) Specification of Gamma-Ray Brachytherapy Sources (1974) Radiological Factors Affecting Decision-Making in a Nuclear Attack (1974)

138 / NCRP PUBLICATIONS 44 Krypton-85 in the Atmosphere-Accumulation, Biological Significance, and Control Technology (1975) 46 Alpha-Emitting Particles in Lungs (1975) 47 Ditium Measurement Techniques (1976) 49 Structural Shielding Design and Evaluation for Medical Use o f X Rays and Gamma Rays of Energies Up to 10 MeV (1976) 50 Environmental Radiation Measurements (1976) 52 Cesium-I37 from the Environment to Man: Metabolism and Dose (1977) 54 Medical Radiation Exposure ofpregnant and Potentially Pregnant Women (1977) 55 Protection of the Thyroid Gland in the Event of Releases of Radioiodine (1977) 57 Instrumentation and Monitoring Methods for Radiation Protection (1978) 58 A Handbook of Radioactivity Measurements Procedures, 2nd ed. (1985) 59 Operational Radiution Safety Program (1978) 60 Physical, Chemical, and Biological Properties of Radiocerium Relevant to Radiation Protection Guidelines (1978) 61 Radiation Safety k i n i n g Criteria for Industrial Radiography (1978) 62 Tritium i n the Environment (1979) 63 Tritium and Other Radionuclide Labeled Organic Compounds Incorporated i n Genetic Material (1979) 64 Influence of Dose and Its Distribution in Time on Dose-Response Relationships for Low-LET Radiations (1980) 65 Management of Persons Accidentally Contaminated with Radionuclides (1980) 67 Radwfrequency Electromagnetic Fields-Properties, Quantities and Units, Biophysical Interaction, and Measurements (1981) 68 Radiation Protection in Pediatric Radiology (1981) 69 Dosimetry of X-Ray and Gamma-Ray Beams for Radiation Therapy in the Energy Range 10 keVto 50 MeV (1981) 70 Nuclear Medicine-Fattors Influencing the Choice and Use of Radionuclides in Diagnosis and Therapy (1982) 71 Operational Radiation Safety-Training (1983) 72 Radiation Protection and Measurement for Low-Voltage Neutron Generators (1983) 73 Protection in Nuclear Medicine and Ultrasound Diagnostic Procedures i n Children (1983) 74 Biological Effects of Ultrasound: Mechanisms and Clinical Implications (1983) 75 Iodine-129: Eoalwrtion of Releases from Nuclear Power Generation (1983) 77 Exposures from the Uranium Series with Emphasis on Radon and Its Daughters (1984)

NCRP PUBLICATIONS /

139

78 Evaluation of Occupational and Environmental Exposures to Radon and Radon Daughters in the United States (1984) 79 Neutron Contamination from Medical Electron Accelerators (1984) 80 Induction of Thyroid Cancer by Ionizing Radiation (1985) 81 Carbon-14 in the Environment (1985) 82 S I Units in Radiation Protection and Measurements (1985) 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 Effectsfrom Internally Deposited Radionuclides (1987) 90 Neptunium: Radiation Protection Guidelines (1988) 92 Public Radiation Exposure from Nuclear Power Generation in the United States (1987) 93 IonizingRadiation Exposure of the Population of the United States (1987) 94 Exposure of the Population i n 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) 98 Guidance on Radiation Received in Space Activities (1989) 99 Quality Assurance for Diagnostic Imaging (1988) 100 Exposure of the U S . Population from Diagnostic Medical 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 uHot Particles" on the Skin (1989) 107 Implementation of the Principle o f A s Low As Reasonably Achievable (ALARQ)for Medical and Dental Personnel (1990) 108 Conceptual Basis for Calculations ofAbsorbed-Dose Distributions (1991) 109 Effects of Ionizing Radiation on Aquatic Organisms (1991)

140 / NCRP PUBLICATIONS 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

113 114 115 116 117 118 119 120 121 122

123 124 125 126 127

for the Assessment of Ionizing Radiation Fields and Radioactive Surface Contamination (1991) Exposure Criteria for Mcdical Diagnostic Ultrasound: I. Criteria Based on Thermal Mechanisms (1992) Maintaining Radiation Protection Records (1992) Risk Estimates for Radiation Protection (1993) Limitation of Exposure to Ionizing Radiation (1993) Research Needs for Radiation Protection (1993) Radiation Protection in the Mineral Extraction Industry (1993) A Practical Guide to the Determination of Human Exposure to Radiofrequency Fields (1993) Dose Control at Nuclear Power Plants (1994) Principles and Application of Collective Dose in Radiation Protection (1995) Use of Personal Monitors to Estimate Effective Dose Equivalent and Effective Dose to Workersfor E x t e m l Exposure to Low-LET Radiation (1995) Screening Models for Releases of Radionuclides to Atmosphere, Surface Water, and Ground (1996) Sources and Magnitude of Occupational and Public Exposures from Nuclear Medicine Procedures (1996) Deposition, Retention and Dosimetry of Inhaled Radioactive Substances (1997) Uncertainties in Fatal Cancer Risk Estimates Used in Radiation Protection (1997) Operational Radiation Safety Program (1998)

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. 32-127).Each binder will accommodate from five to seven reports. The binders cany 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 11. NCRP Reports Nos. 23,25,27,30 Volume 111. NCRP Reports Nos. 32, 35,36, 37 Volume W. NCRP Reports Nos. 38,40,41 Volume V. NCRP Reports Nos. 42,44,46 Volume VI. NCRP Reports Nos. 47,49,50,51 Volume MI. NCRP Reports Nos. 52,53,54,55,57

NCRP PUBLICATIONS / 141 Volume VIII. NCRP Report No. 58 Volume M.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 XM. 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 XXnr. NCRP Reports Nos. 119,120,121,122 Volume MCV. NCRP Report No. 1231 and 12311 (Titles of t h e individual reports contained in each volume a r e given above.)

NCRP Commentaries

Title

No.

Krypton-85 in the Atmosphere-With spec if^ Reference to the Public Health Significance of the Proposed Controlled Release at Three Mile Island (1980) Screening Techniques for Determining Compliance with Environmental Standards-Releases of Radionuclides to the Atmosphere (1986).Revised (1989) Guidelines for the Release of Waste Water @ m 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 . PopulationStatus of the Problem

(1991) Misadministration of Radioactive Material i n Medicine-Scientific Background (1991) Uncertainty in NCRP Screening Models Relating to Atmospheric Dansport, Deposition and Uptake by Humans (1993) Considerations Regarding the Unintended Radiation Exposure of the Embryo, Fetus or Nursing Child (1994)

142 / NCRP PUBLICATIONS 10 Advising the Public about Radiation Emergencies:A Document for Public Comment (1994) 11 Dose Limits for Individuals Who Receive Exposure from Radionuclide Therapy Patients (1995) 12 Radiation Exposure and High-Altitude Flight (1995) 13 An Introduction to Efficacy in Diagnostic Radiology and Nuclear Medicine (Justification of Medical Radiation Exposure) (1995) 14 A Guide for Uncertainty Analysis in Dose and Risk Assessments Related to Environmental Contamination (1996)

Proceedings of the Annual Meeting No.

Title

Perceptions of Risk, Proceedings of the Fifteenth Annual Meeting held on March 14-15, 1979 (including Taylor Lecture No. 3) (1980) 3 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) 4 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) 5 Environmental Radioactivity, Proceedings of the Nineteenth Annual Meeting held on April 6-7, 1983 (including Taylor Lecture No. 7) (1983) 6 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) 7 Radioactive Waste, Proceedings of the Twenty-first Annual Meeting held on April 3-4, 1985 (including Taylor Lecture No. 9)(1986) 8 Nonionizing Electromagnetic Radiations and Ultrasound, Proceedings of the Twenty-second Annual Meeting held on April 2-3, 1986 (including Taylor Lecture No. 10) (1988) 9 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) 10 Radon, Proceedings of the Twenty-fourth Annual Meeting held on March 30-31, 1988 (including Taylor Lecture No. 12) (1989) 11 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) 1

NCRP PUBLICATIONS /

143

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 i n 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) 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 ofNew Data on Radiation Cancer Risk, Proceedings o f the Thirty-second Annual Meeting held on April 3-4,1996 (including Taylor Lecture No. 20) (1997)

Lauriston S. Taylor Lectures No. 1

2 3 4

Title The Squares of the Natural Numbers i n Radiation Protection by Herbert M. Parker (1977) Why be Quantitative about Radiation Risk Estimates? by Sir Edward Pochin (1978) Radiation Protection--Concepts and n u d e Offs by Hymer L. Friedell (1979) [Available also in Perceptions of Risk, see abovel From "Quantity of Radiation" and "Dose" to "Exposure" and "Absorbed Dosen-An Historical Review by Harold 0.Wyckoff (1980) How Well Can We Assess Genetic Risk? Not Very by James F.Crow (1981) [Available also in Critical Issues i n Setting Radiation Dose Limits, see above] Ethics, Dude-offs and Medical Radiation by Eugene L. Saenger (1982) [Available also in Radiation Protection and New Medical Diagnostic Approaches, see abovel The Human Environment-Past, Present and Future by Merril Eisenbud (1983) [Available also in Environmental Radioactivity, see abovel Limitation and Assessment i n Radiation Protection by Harald H. Rossi (1984) [Available also in Some Issues Important i n Developing Basic Radiation Protection Recommendations, see above]

144 / NCRP PUBLICATIONS Truth (and Beauty) in Radiation Measurement by John H . Harley (1985) [Available also in Radioactive Waste, see abovel Biological Effects of Non-ionizing Radiations: Cellular Properties and Interactions by Herman I? Schwan (1987)[Available also in Nonionizing Electromagnetic Radiations and Ultrasound, see above] 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 Rudiation Protection: The Past Century and Prospects for the Future by Arthur C. Upton (1989) [Available also in Radiation Protection Today, see abovel Radiation Protection and the Internal Emitter Saga by J. Newel1 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 abovel 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)

Symposium Proceedings

No. 1

2

3

Title 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)

NCRP PUBLICATIONS / 145

NCRP Statements No.

Title

1

"Blood Counts, Statement of the National Committee on Radiation Protection," Radiology 63,428 (1954)

2

"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)

3

X-Ray Protection Standards for Home levi is ion Receivers, Interim Statement of the National Council on Radiation Protection and Measurements (1968)

4

Specifiation of Units ofNatural Uranium and Natural Thorium, Statement of the National Council on Radiation Protection and Measurements (1973)

5

NCRP Statement on Dose Limit for Neutrons (1980)

6

Control of Air Emissions of Radionuclides (1984)

7

The Probability That a Particular Malignancy May Have Been Caused by a Specified Irradiation (1992)

Other Documents The following documents of the NCRP 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, February 19,1960, Vol. 131, No. 3399, pages 482-486 Dose Effect Modihing 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)

Index Access controls 45-47, 59 audible signals 46 barriers 46 external exposure controls 45 interlock 46 internal exposure controls 59 pre-startup notification 46 run-safe switches 46 visual signals 46 Access control systems 47 effective dose category 47 Accident investigations 21 Administrative controls 6,43, 49,57 Air monitoring 56.66-69 criteria for surveys 67 internal dose assessment 68 objectives 68 ALARA (as low as reasonably achievable) 2,3,5-11, 12,18, 26,40,43-44, 52,55, 61,81 administrative controls 6 applicability of cost-benefit analysis 7 collective effective dose 11 cost-benefit approach 8 documentation 11 external exposure controls 43-44, 52 facility design 26 management commitment and policy 12 monetary value of avoided dose 9-10 occupational dose records 55 occupational exposures 11 public exposure 11,81 screening for ALAFtA assessment 11

social and economic factors 7 use of collective effective dose 8 Alarm systems 45-47,59 area monitors 46 audible signals 46 criticality accident 47 effective dose category 47 external exposure controls 45 internal exposure controls 59 pre-startup notification 46 visual signals 46 Area monitoring 53 external exposure controls 53 Area surveys 19 Audits 20,22 Bioassay 56, 62-66, 71 direct bioassay 63-65 dose assessment 65,66 indirect bioassay 64,65 Calibration 98 . facility 98 frequency 98 Collective effective dose 8

ALARA 8 Contamination controls 56 Contamination control programs 57-59 engineering controls 58 Contamination surveys 69 Cost-benefit analysis 7 ALARA 7 Criticality accident 51 personal monitoring 51 Criticality safety 36 facility design 36 Decontamination of equipment 76 waste 76 Detriment 9 social factors 9

INDEX /

stochastic health effects 9 Dose records 54,71,72 external exposure controls 54 internal exposure controls 71, 72 Emergency plan 108-113 development 108,109 equipment 110 implementing procedures 109 key individuals 109,110 plan evaluation 112,113 Emergency procedures 17 Emergency response 22 Engineering controls 42,43,49, 58,61 ALARA 61 external exposure controls 43 internal exposure controls 58 Environmental monitoring 82-93 documentation 92,93 dose assessment 91-92 measurement methods 90-91 operational monitoring 89-90 preoperational monitoring 88-89,90 purposes 87,88 quality assurance 92-93 screening models 88 Exposure limits 4 External dosimetry 17 External exposure controls 42-55 access controls 45-47 access control systems 46-47 administrative dose guidelines 43 ALARA 43,44 alarm systems 45-47 area monitoring 53 dose assessment 51 dose records 54 engineering controls 43,49 facility design 42 personal dosimetry 42 personal monitoring 49-51

147

protective clothing 53 radiation safety procedures 48-49 radiation surveys 51 radiation work permits (RWP) 43,48-49 shielding 44 Facility design 26-37.42 air cleaning devices 33 criticality safety 36 economic considerations 28 equipment and system design 29

exhaust vents 33 facility layout 28 fume hoods 34 gloved boxes 34 instrumentation and access control systems 36 monitoring and surveillance equipment 29-30 personal decontamination 29 power reactors 27 public or commercial access 28 radiation-producing equipment 27 radioactive material 27 radioactive waste management 35 shielding 30 site selection 26-28 ventilation 32 waste management 29 waste processing facilities 27 Fume hoods 34-35 Glwed boxes 3435,41 training 41 Incident investigations 21 Instrumentation 94-107 accuracy acceptance criteria 101 calibration 96-98 etchable track detectors 105 film 105

148 / INDEX Geiger-Miiller counters 102, 104,105 instrument specification 95 ionization chamber 102,104, 105 liquid drop (bubble) detectors 104 maintenance 98-99 measurement accuracy 99 measurement uncertainty 100 pulse counting instruments 102 radioactive contamination 103 records 106,107 scintillation detectors 102,105 selection of instrument for various applications 100 semiconductor surface barrier 105 short-term stability 101 superheated liquid drop (bubble) detectors 105 thermoluminescent detectors (TLD)104 uncertainties 97 use categories 99 Interlocks 43 Internal dosimetry 17 Internal exposure controls 56-72 air monitoring 56 access controls 59 alarm systems 59 bioassay 56,62-65 contamination controls 56 contamination control programs 57-59 dose assessments 56,65-66, 68-69 dose records 71-72 engineering controls 58,61 limits 57 monitoring and surveillance program 66-69 personal monitoring 56,62,65 protective clothing 70 protective equipment and devices 69-71

radiation safety procedures 61 radiation safety program 60 radiation work permits (RWP) 60-61 respiratory protection 56,70 Limitation 5 Limits 4,30,43,57 external exposure controls 43 facility design 30 internal exposure controls 57 summary of limits 4 Justification 5 Management commitment and policy 12-13 ALARA 12 budgetary support 12 goals of the radiation safety program 12 Radiation Safety Officer (RSO) 13 worker training 13 Monetary value of avoided dose 9,lO ALARA 9.10 Monitoring and surveillance program 66-69,83-90 airborne radioactivity 66-69, 84-85 emergency 83-84.90 environmental 87-90 internal dose assessment 68-69 liquid effluent 86 solid waste 86-87 Negligible individual dose 4,9 Occupational dose records 55 ALARA 55 Occupational exposure 4,42 limit 4 Occupational medicine 23 Off-site exposures 82-86 airborne effluents 84-85

INDEX /

determining the need for monitoring 83 liquid effluents 86 sampling frequency 85 screening models 83 solid waste 86 techniques for control 82 Operating procedures 42

149

goals 18 incident investigations 21 personal monitoring 19 program audits 20 surveillance 19 surveys 19,20 Radiation emergencies 40,41,

108-113 classification of emergencies

110-111 Personal decontamination 29 Personal dosimetry 42,49 Personal monitoring 19,22,

49-51,56,62-65 bioassay 62-65 criticality accident 51 dose assessment 51 dosimeter devices 50 external radiation dosimetry

49 internal exposure controls 62 radiation surveys 51 Protective clothing 17,42,53,70 Protective equipment and devices 69-71 containment systems 69 internal exposure controls

70-71 protective clothing 70 respiratory protection 70 Public exposure 4,81-93 environmental monitoring 87,

88 limit 4 solid waste 87 standards 81 Quality assurance 18-22,92-93 accident investigations 21 area surveys 19 audits 22 corrective actions 22 deficiency tracking 22 definition 18 environmental monitoring

92-93

emergency plan 108 practical considerations 111 restoration and recovery 111 training 40,41 Radiation Safety Committee (RSC) 13-16,18,22 Radiation safety manual 15-16 Radiation Safety Officer (RSO)

13,14,15,18,19,21,22,23 authority 14 qualification 14 responsibility 14 Radiation safety procedures 15,

16-17,48-49,60-61 Radiation surveys 17.51-52 documentation 52 nonroutine 52 routine 52 shielding 52 Radiation work permits (RWP)

43,48-49,60,61 Radiological accident 33 Records 54-55,71-72,79,93,

106-107 environmental monitoring 93 external exposures 54-55 instrumentation 106-107 internal exposures 71-72 waste 79 Records management 23 Respiratory protection 17,41,

56,67,70 Respiratory protective equipment 32.41 training 41 Responsibilities 17-19 management 18-19

150 /

INDEX

Radiation Safety Committee (RSC) 18 Radiation Safety Officer (RSO)

18-19 worker 17 Shielding 30-32,52 dose limit 30 facility design 30 materials 31 radiation surveys 52 Site selection 26-27 facility design 27 power reactors 27 radiation-producing equipment 27 radioactive material 27 waste processing facilities 27 Surveys 19,22 Training 22,38-41,75,112-113 design of a general training Program 39 general principles 38 gloved box 41 radiation emergencies 40-41,

112-113 respiratory protection equipment 41 specific training requirements

41 waste minimization 75 Ventilation 32-35 air cleaning devices 33 exhaust vents 33 facility design 32 radiological accident 33 Waste 73-79,86 classifying 76-77 collecting 76-77 disposal 78 minimizing 74-76 mixed wastes 73,75-76 records 79,86 recycling 79

storage 78 sorting 76-77 volume reduction 77 Waste management 22,29 Worker qualifications 17