Tài liệu Ncrp report no 127 operational radiation safety program

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-Publication Data 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 11 Purpose of this Report . . . . . . . . . . . . . . . . . . . . . . . . . . 1 . 12 Purpose of the Operational Radiation Safety . Program ..................................... 2 2 Application of ALARA . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 21 Applicability of Cost-Benefit Analysis in the . ALARA Process ............................... 7 22 Concepts of a Cost-Benefit Approach to ALARA . . . . . 8 . 2 2 1 Applicability of Collective Effective Dose . . . . . 8 .. 2 2 Dose Magnitude and Distributions . . . . . . . . . . 8 .3 223 Monetary Value of Avoided Dose ........... 9 .. 23 Screening for ALARA Assessment . . . . . . . . . . . . . . . 11 . 3 Organization and Administration . . . . . . . . . . . . . . . . . 12 31 Management Commitment and Policy . . . . . . . . . . . . 12 .. 32 Radiation Safety Organization . . . . . . . . . . . . . . . . . . 13 . 321 Radiation Safety Advisory Organization . . . . 13 .. 3 2 2 Radiation Safety Officer . . . . . . . . . . . . . . . . . 14 .. 33 Accreditation and Certification ................. 14 . 3.4 Radiation Safety Program Policies and Procedures .................................. 15 341 Radiation Safety Manual ................. 15 .. 342 Radiation Safety Operating Procedures . . . . . 16 .. 35 Responsibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 . 36 Quality Assurance ............................ 18 . 361 Management Goals ..................... 19 .. 3 6 2 Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 .. 363 Program Audits ........................ 20 .. 364 Incident and Accident Investigations ....... 21 .. 365 Deficiency Tracking . . . . . . . . . . . . . . . . . . . . . 22 .. 37 Records Management ......................... . 23 38 Occupational Medicine ........................ 23 . 39 Recommended Additional Reading ............... 24 . 4 Facility Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 41 Site Selection ................................ 26 . 42 Facility Layout ...............................28 . . . . . vii 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 . . . 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 . x / CONTENTS 1 . Selection of Instruments for Various Applications . 100 05 1 . Records for an Instrument Program . . . . . . . . . . . . .106 06 1 . Recommended Additional Reading . . . . . . . . . . . . . . 107 07 11 Planning for Radiation Emergencies . . . . . . . . . . . . .108 1 . Development of the Emergency Plan . . . . . . . . . . . .108 11 1 . Preparation of Implementing Procedures . . . . . . . . . 109 12 1 . Classification of Emergencies . . . . . . . . . . . . . . . . . .110 13 11.4 Practical Considerations . . . . . . . . . . . . . . . . . . . . . . 111 11.5 Evaluation of the Plan . . . . . . . . . . . . . . . . . . . . . . . . 112 Glossary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .114 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 TheNCRP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 NCRPPublications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 137 Index . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146 . 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 / 3 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 TABLE .l-Summary of NCRP recommendations specifying 1 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 x 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 6 / 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. 8 1 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 1 9 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 10 / 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 tive effective dose by an increment (AS) t a cost not exceeding the 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.