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
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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
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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ionizing radiation (NCRP Report No. 91 in 1987 which was superseded by NCRP Report No. 116 in 1993); new techniques for the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
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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.
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