Radiation Protection in the Mineral Extraction Industry: Recommendations of the National Council on Radiation Protection and Measurements (Ncrp, No)

NCRP REPORT No. 118

RADIATION PROTECTION IN THE MINERAL EXTRACTION INDUSTRY Recommendations of the NATIONAL COUNCIL O N RADIATION PROTECTION AND MEASUREMENTS

Issued November 30,1993

National Council on Radiation Protection and Measurements 7910 Woodmont Avenue 1 Bethesda, MD 20814-3095

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

Library of Congress Cataloging-in-Publication Data National Council on Radiation Protection and Measurements. Radiation protection in the mineral extraction industry : recommendations of the National Council on Radiation Protection and Measurements. cm.-(NCRP report ; no. 118) p. "Prepared by Scientific Committee 46-2 on Uranium Mining and MillingRadiation Safety Programs"-Pref. "Issued November 30, 1993." Includes bibliographical references (p. ) and index. ISBN 0-929600-33-9 1. Mine safety. 2. Ore-dressing plants-Safety measures. 3. RadiationSafety measures. I. National Council on Radiation Protection and Measurements, Scientific Committee 46-2 on Uranium Mining and MillingRadiation Safety Programs. IT. Title. 111. Series. TN295.N28 1993 622l.8-dc20 93-33554 CIP

Copyright O National Council on Radiation Protection and Measurements 1993 All rights reserved. This publication is protected by copyright. No part of this publication may be reproduced in any form or by any means, including photocopying, or utilized by any information storage and retrieval system without written permission from the copyright owner, except for brief quotation in critical articles or reviews.

Preface This Report was originally intended as radiation protection recommendations for the uranium mining and milling industry. The Committee early on, however, recognized t h a t there were known radiation problems connected with the mining and milling of several minerals. Further, the Committee recognized that the extraction and processing of virtually any mineral might result in some level of radiation exposure and that the application of radiation protection practices may be warranted in some cases. Therefore, the Report that evolved addresses the whole mineral industry and the material prepared for the uranium mining and milling industry was retained to provide examples of the more complex problems encountered and solutions to those problems. The Report was written for a specific audience-management and its technical staff-who either have been made aware of radiation problems by the imposition of a regulation or perceive the importance of evaluating facility design and operations in the societal context of a greater awareness about occupational and environmental risks. The International System of Units (SI) is used herein and, in accordance with the recommendations set forth in NCRP Report No. 82, SI Units in Radiation Protection and Measurements, the use of conventional units has been discontinued. Appendix B contains a conversion table of SI units and conventional units. This Report was prepared by Scientific Committee 46-2 on Uranium Mining and Milling-Radiation Safety Programs, working under the auspices of Scientific Committee 46 on Operational Radiation Safety. Serving on Scientific Committee 46-2 were:

Richard L. Doty, Chairman Pennsylvania Power and Light Company Allentown, Pennsylvania Members Albert J. Hazle Arvada, Colorado

Noel Savignac Albuquerque, New Mexico

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PREFACE

Charles E. Roessler University of Florida Gainesville, Florida

Edwin T. Still Kerr-McGee Corporation Oklahoma City, Oklahoma

Scientific Committee 46 Liaison Member

Keith Schiager University of Utah Salt Lake City, Utah Serving on Scientific Committee 46 were:

Kenneth R. Kase, Chairman (1990Stanford Linear Accelerator Center Stanford, California

)

Charles B. Meinhold, Chairman (1983-1990) Brookhaven National Laboratory Upton, New York Members

Ernest A. Belvin (1983-1987) Marietta, Georgia

David S. Myers (1987Lawrence Livermore National Laboratory Livermore, California

W. Robert Casey (1983-1989) Brookhaven National Laboratory Upton, New York

John W. Poston, Sr. (1991- ) Texas A&M University College Station, Texas

Robert J. Catlin (1983-1992) University of Texas Houston, Texas

Keith Schiager (1983University of Utah Salt Lake City, Utah

Joyce P. Davis (1990- ) Defense Nuclear Facilities Safety Board Washington, D.C. William R. Hendee (1983- ) Medical College of Wisconsin Milwaukee, Wisconsin

)

)

Ralph H. Thomas (1989- ) Lawrence Livermore National Laboratory Livermore, California Robert G.Wissink (1983- ) Minnesota Mining and Manufacturing (3M) Center St. Paul, Minnesota

PREFACE

James E. McLaughlin (1983- ) Santa Fe, New Mexico

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Paul L. Ziemer (1983-1990) U.S. Department of Energy Washington, D.C.

Thomas D. Murphy (1983-1992) U.S. Nuclear Regulatory Commission Washington, D.C. NCRP Secretariat

James A. Spahn, Jr. (1986- ) Robert T. Wangemann (1986) E. Ivan White (1983-1985) The Council wishes to express its appreciation to the members of the Committee for the time and effort devoted to the preparation of this Report. Charles B. Meinhold President, NCRP Bethesda, Maryland August 1,1993

Contents Preface ....................................................................................... 1 Introduction ......................................................................... 1.1 Purpose ............................................................................ 1.2 Concepts of Radiation Protection .................................. 1.3 Scope of Report ............................................................... 2 Design of Radiation Protection Programs ................... 2.1 Criteria for Radiation Protection Programs ................. 2.2 Program Management ................................................... 2.3 Radiation Safety Officer ................................................. 2.4 Radiation Safety Committee ......................................... 2.5 Preparation and Maintenance of Records .................... 2.6 Quality Assurance Program ........................................ 2.7 Coordination Among Safety Programs ......................... 3 Sources of Potential Radiation Exposures ................... 3.1 Source Characterization ................................................. 3.1.1 Naturally Occurring Radioactive Materials ...... 3.1.2 Distribution of Radioactivity in Ore. Product, By-Products and Wastes ................................... 3.1.3 Characteristics Related to Radiation Dose ........ 3.2 Occupational Exposures ................................................. 3.2.1 External Radiation ............................................... 3.2.2 Airborne Radioactivity ....................................... 3.2.3 Surface Contamination ........................................ 3.3 Releases to the Environment ........................................ 3.3.1 Airborne ................................................................ 3.3.2 Waterborne ............................................................ 3.3.3 External Radiation ............................................... 3.4 Process By-Products and Waste Materials ................... 4 Exposure Management Program .................................... 4.1 Exposure Limits ............................................................. 4.2 Exposure Environment .................................................. 4.2.1 External Radiation ............................................... 4.2.2 Ore Dust ................................................................ 4.2.3 Airborne Radon and Radon Progeny .................. 4.3 Facility Design and Engineering .................................. 4.3.1 Site Selection ........................................................ 4.3.2 Facility Layout ..................................................... 4.3.3 Equipment and System Design ...........................

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CONTENTS

4.4 Facility Procedures and Practices ................................. 4.4.1 Access Control ...................................................... 4.4.2 Radioactive Material Control .............................. 4.4.2.1 Materials Handling ................................. 4.4.2.2 Waste Management ................................. 4.4.2.3 Sealed Source Control ............................. 4.4.3 Personnel Protective Equipment ........................ 4.4.3.1 Respiratory Protection ............................ 4.4.3.2 Protective Clothing ................................. 4.5 Employee Training .......................................................

5. Monitoring of Occupational Exposure ................................. 5.1 Monitoring Objectives .................................................... 5.1.1 Characterization of the Workplace ..................... 5.1.2 Personnel Exposure Assessment ......................... 5.2 Monitoring Program ....................................................... 5.3 External Radiation ......................................................... 5.3.1 Characterization of the Workplace ..................... 5.3.2 Personal Monitoring-External .......................... 5.4 Long-Lived Airborne Radionuclides ............................. 5.4.1 Characterization of the Workplace ..................... ....... 5.4.2 Personnel Exposure Assessment-Internal 5.5 Airborne Radon and Progeny ....................................... 5.5.1 Characterization of the Workpla~e-~~~Rn and

Progeny

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5.5.2 Personnel Exposure Asse~sment-~~~Rn and

Progeny

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5.5.3 Radon-220 and Progeny ....................................... 5.5.4 Monitoring for Control Purposes ........................

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5.6 Surface Contamination .................................................. 5.6.1 Area Monitoring ................................................... 5.6.2 Monitoring of Personnel ...................................... 5.6.3 Monitoring Other Items ....................................... 5.7 Bioassay .......................................................................... 5.7.1 Bioassay Methods ............................................... 5.7.2 Bioassay Program Content .................................. 5.7.3 Routine Bioassay .................................................. 5.7.4 Post-Exposure and Follow-Up Measurements ...

6 Effluent Monitoring and Environmental

Surveillance

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6.1 Environmental Pathways .............................................. 6.2 Effluent Monitoring ........................................................ 6.2.1 Effluent Monitoring Objectives ........................... 6.2.2 Program Design ................................................... 6.2.3 Air Monitoring ..................................................... 6.2.4 Water Monitoring .................................................

CONTENTS

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6.3 Environmental Surveillance .......................................... 6.3.1 Environmental Monitoring Objectives ............... 6.3.2 Program Design .................................................... 6.3.3 Radon .................................................................... 6.3.4 Radon Progeny ..................................................... 6.3.5 Long-Lived Airborne Radionuclides ................... 6.3.6 Soil and Vegetation .............................................. 6.3.7 Water ..................................................................... 6.3.8 External Radiation ...............................................

7 Guidelines. Standards and Regulations ........................ 7.1 General ............................................................................ 7.2 Sources of Guidance and Standards .............................. 7.2.1 Scientific Recommendations ................................ 7.2.2 Consensus Standards ........................................... 7.2.3 Federal Guidance and Policy .............................. 7.2.4 Rules and Regulations ......................................... 7.3 Approaches to Radiation Limits .................................... 7.4 Occupational Exposures ................................................. 7.4.1 Introduction .......................................................... 7.4.2 Recommendations ................................................. 7.4.3 Standards and Regulations ................................. 7.5 Effluents and the Environment ..................................... 7.5.1 Emuents ................................................................ 7.5.2 Wastes ................................................................... 7.5.3 Uranium and Thorium Processing Sites ............ 7.5.4 Other ..................................................................... 8 Radiation Emergency Response Planning ................... 8.1 General ............................................................................ 8.2 Operations ....................................................................... 8.3 Environment ................................................................... 8.4 Transportation ................................................................ 9 Radiation Protection in Specific Applications ............ 9.1 Introduction .................................................................... 9.2 Heap-Leach Extraction .................................................. 9.3 I n situ Mineral Extraction ............................................. 9.4 Side-Stream Extractions of Uranium ........................... 9.4.1 Uranium Recovery from Phosphoric Acid .......... 9.4.2 Occupational Exposure Considerations .............. 9.4.3 Shipping and Transportation .............................. 9.4.4 Efluents and Environmental Monitoring .......... 9.4.5 Solid Radioactive Waste and Equipment Reuse

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or Salvage

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9.5 Thorium and Rare-Earths Processing .......................... 9.6 Phosphate ........................................................................

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CONTENTS 9.6.1 Mining. Beneficiation and Wet Rock

Handling

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9.6.1.1 Occupational Exposure ........................... 9.6.1.2 Mining and Beneficiation Wastes and

Post-Mining Land

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9.6.1.3 Liquid Releases and Water Quality 9.6.2 Phosphate Rock Drying and Dry Rock

Handling

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9.6.2.1 Occupational Exposure 9.6.2.2 Emissions

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9.6.3 Wet-Process Phosphoric Acid Plants ................... 9.6.3.1 Occupational Exposure-Protection

Operations

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9.6.3.2 Occupational Exposure-Clean-up

and Maintenance ........................................ 9.6.3.3 Occupational Exposure-Filter Pan Repair ................................................... 9.6.3.4 Waste Management ................................. 9.6.3.5 Phosphogypsum ....................................... 9.6.4 Production of Phosphate Products ...................... 9.6.5 Thermal Process (Elemental Phosphorus) ......... Appendix A Radioactive Serie~.~~'U. and ='U ...... Appendix B Conversion Factors ........................................ Glossary ..................................................................................... References ................................................................................. The NCRP ................................................................................. NCRP Publications ................................................................. Index ...........................................................................................

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1. Introduction 1.1 Purpose

The National Council on Radiation Protection and Measurements (NCRP) develops recommendations dealing with various aspects of operational radiation protection. The basic principles and practices of radiation safety are well established. However, specific facilities present specific problems. To serve the needs of a particular facility, effective programs should recognize and account for variables such as the complexity of radiation exposure pathways and the magnitude of potential radiation exposure a t a given facility. This Report describes the vital parts of an effective radiation safety program for mineral extraction facilities. It provides information useful for choosing appropriate techniques of radiation control and monitoring a t such facilities. Because radioactive material occurs naturally throughout the earth's crust, any mineral extraction operation or process, not just those commonly perceived to be processing radioactive materials, is a candidate for radiation safety measures. This Report draws on examples from the uranium mining and milling industry, but principles and practices common across the entire mineral extraction industry are emphasized. Mining, milling and beneficiation have long been accepted technologies for extracting and processing ores. However, they are among the technologies that have come under increasing scrutiny from a society concerned about occupational and environmental risks. Increasing awareness and attention has been placed on the potential uses and risks of radioactive materials. Therefore, assessing the radiation protection requirements and practices of the mineral extraction industry is both timely and consistent with good work practices. This Report is written so that individuals with a basic technical background can apply the concepts of radiation protection to evaluate any mineral extraction operation. Management can use the Report to define the degree to which radiation safety should be considered in designing facilities and planning their operation. Design engineers as well a s health and safety professionals will find useful

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INTRODUCTION

information for applying the basic principles and practices of radiation safety to their specific facility's design and program. The reader is not presumed to be thoroughly familiar with other radiation safety literature. This Report addresses primary aspects of radiationprotection, but no single document can provide the solution to all problems which may arise relating to radiation safety. References are provided for those who wish to obtain more detailed information on specific topics.

1.2 Concepts of Radiation Protection Radiation protection programs are designed to allow society to gain the benefits of using radioactive materials while minimizing risk to the public and workers. In the case of mineral extraction, the goal is to make sure that minerals can be extracted and processed while keeping risks from radiation exposure to a minimum. The key is to make sure exposures are evaluated and controlled effectively. Most situations involving exposure to radioactive materials can be managed easily because exposure to personnel can readily be maintained well within established exposure limits; the radiation risk is controlled using simple and inexpensive techniques. As exposure to personnel approaches the limit, however, more technically complex and expensive controls need to be considered. One goal of the radiation safety program is to make sure that both users and the public are protected a t reasonable cost. Adequate protection means that risk should, as far as possible, be limited to levels comparable with those experienced in other safe industries (NCRP, 1993). The basis for controlling radiation-induced risk is drawn from the many scientific studies which have been performed and which have resulted in adoption of what is called the "linear no-threshold hypothesis." In simple terms, the assumption is that for any increase in exposure, there is a corresponding increase in risk. Applying the theory to real-life situations has its complications. Biological effects that can be produced by radiation are also caused by other physical and chemical agents and also occur naturally. Above-normal incidence of these effects (e.g.,several types of cancer) has been observed in individuals exposed a t radiation levels greatly in excess of those discussed in this Report as individual exposure "limits." To assess the more common exposure situations, scientists extrapolate from the number of observed effects a t high exposures to predict the number of presumed effects at lower exposures. The difficulty arises primarily because, a t low exposures, there is a very low probability

1.3 SCOPE OF REPORT

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that radiation-induced effects will occur. Further, any effects that do occur are indistinguishable from those induced by other agents and those normally seen frequently in any population. The assumption is made that effects will result from exposure of a population to radiation and that the number of people affected, or the risk to a specific individual, will be in direct proportion to the total radiation exposure. This cautious assumption may overestimate risk but has been adopted by the NCRP for purposes of radiation protection. Because any radiation exposure is assumed to have an associated risk, exposures are to be maintained at levels which are as low as reasonably achievable (ALARA), economic and social factors being taken into account. The inclusion of the word "reasonably" recognizes both that the use of radioactive materials can yield benefits to society and that exposure reduction often requires resource expenditures. Achievement of exposure levels which are ALARA reflects the application of exposure reduction techniques until further reduction can be attained only if the intended benefit would not be obtained or the cost would be unreasonable. Exposure management to levels as low as reasonably achievable is accomplished by controlling a number of variables: the quantity of radioactive material, the location of workers and the public relative to the material, the length of time people are exposed to the material, the amount of material that inadvertently escapes from processing streams, and the effluents that contain radionuclides and are released to the environment. An effective radiation safety program thus includes evaluation of situations which may lead to radiation exposure, comparison of expected exposures to exposure limits mandated by regulation, and the application of control practices to maintain exposure at levels which are ALARA.

1.3 Scope of Report

This Report describes the application of radiation safety concepts to mining, milling and beneficiation facilities. The potential for radiation exposure differs in magnitude depending on the mineral and the facility type. However, the same basic principles of radiation safety should be applied to facility design and operations. For some facilities, the application of these principles may mean the establishment of a radiation safety program because radiological protection had not previously been considered relevant to the operation. Therefore, this Report emphasizes fundamental concepts, simplifiestechnical terminology and presents methods based on past experiences

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1. INTRODUCTION

with mineral extraction. This Report will deal primarily with design and operation. The decommissioning of facilities and reclaiming of the facility site are beyond the scope of this Report. Each section of this Report presents a piece of the total picture on how to design and operate an effective radiation protection program. In Section 2, organizational structures applicable to different program needs are discussed; while in Section 3, the characterization of radiation sources and the potential pathways to individual exposure are addressed. In Section 4, management of radiation exposures by facility design and in facility operations is described. Concepts of monitoring occupational radiation exposure are considered in Section 5, and the monitoring of potential exposures to the public around a facility are described in Section 6. In Section 7, pertinent regulations, standards and guidelines are presented. Emergency planning concepts are addressed in Section 8. In Sections2 through 8, concepts applicable to any mineral extraction facility are illustrated, and guidelines are provided for program implementation. Uranium extraction is specifically discussed to illustrate program elements which can be applied to the exposure potential of specific mineral extraction processes and to direct the reader to an appropriate level of radiological control for those processes. In the last section of the Report, Section 9, information is presented about radiation safety practices for some specific extraction processes, such as phosphate mining and processing operations.

2. Design of Radiation protection Programs 2.1 Criteria for Radiation Protection Programs Managers of mineral extraction facilities are responsible for ensuring that health, life, property and the environment are protected during the conduct of operations. This requires a knowledge of the particular types and levels of risk associated with their facilities. In some cases, the predominant risk may be from common industrial hazards; in others, from toxic materials. When exposure to radiation or radioactive materials contributes to risk, consideration must be given to radiation safety measures necessary to protect the health and safety of employees and the general public. The motivation to develop an effective radiation safety program can come from several sources. For some processes, regulations mandate evaluating risks associated with naturally radioactive materials in ore and waste products. Other extraction processes are unregulated (for purposes of radiation protection) but may result in radiation exposure as ores are brought to the surface and processed. Other motivators include corporate policy, the societal trend toward litigation where exposure to a risk-producing agent is involved, and, most important, concern for the health and welfare of both employees and the public. Whatever the motivation, there is a need for making responsible decisions about the level of radiological control appropriate to a given facility. In the process of developing a specific radiation safety program, each of these motivators should be considered. The goal is that outlined in Section 1.2: Evaluating facility operations with the aim of reducing exposures to levels which are within established limits and ALARA. If this intent is met, several benefits may result, includingreduced risk to individuals, improved worker morale, enhanced public and1 or regulatory agency perception of the facility, and reduced vulnerability to workmen's compensation claims or radiation-related litigation. For some facilities, exposure may be routinely so small that an ongoing radiation safety program is unnecessary. Assessments such

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2. DESIGN OF RADIATION PROTECTION PROGRAMS

as those described by Dixon (1984)may be useful in identifying those facilities or processes for which the need for ongoing radiation safety practices should be investigated further. As exposure to radiation or radioactive materials becomes more frequent and as radiation levels become more significant, the application of appropriate levels of radiation protection controls is warranted. Based on the statement in Section 1.2, the radiation safety program should be designed to limit risks to employees and members of the public to levels comparable with risks from other common contributors to risk. For example, the level of safety provided for employees should ensure an average risk from radiation no greater than that from all sources of risk for workers in "safe" industries (NCRP, 1993).

2.2 Program Management Radiation safety programs will vary in staffing and structure, depending on the degree of potential radiation exposure which has been identified and the anticipated difficulty of controlling it. This Section of the Report describes a decision chain useful for setting up a radiation safety program appropriate to the needs of a given facility. Staffing demands will vary with the severity of the conditions a t the facility. For example, if the projected exposure for any individual is well below the annual exposure limit for a member of the public, employing radiation specialists may be unnecessary. When t h e potential exposure may approach or exceed this annual limit, consideration should be given to utilizing corporate staff or consultants to evaluate conditions. For those cases where potential exposures to some individuals are anticipated to exceed the limits for members of the public, staffing becomes a choice between a single organization responsible for both industrial and radiological health and safety, or a qualified staff dedicated solely to radiation protection. At some facilities the first choice may be sufficient. At facilities requiring a more extensive occupational (Section 5) or environmental (Section 6) monitoring program, a dedicated staff may be warranted. Finally, where radiation exposure pathways are more complex or difficultto control, the facility may need to designate a radiation safety officer and appoint a radiation safety committee to review radiation safety issues and recommend actions to senior management (NCRP, 1978a).That level of need should rarely occur in the mineral extraction industry, and when i t does, quantities of radioactive materials and potential

2.2 PROGRAMMANAGEMENT

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exposures are likely to be large enough that some program structure may be imposed by regulatory agencies. Regardless of the size and configuration of a facility's radiation safety organization, management retains the responsibility for maintaining exposures within limits and ALARA. The safety organization's role is to provide technical support and equipment, so that radiation protection practices and the concept of ALARA can be applied most effectively. Optimal collaboration between management and the radiation organization must begin early. When radiation safety evaluations and criteria are applied as soon as possible in facility planning, the facility is more likely to be designed and operated in ways consistent with the objective of limiting exposures in practical, cost-effective ways. When the size of potential radiation exposures requires development of ongoing controls, management's responsibility may extend for some period of time, even to the point of providing maintenance and monitoring programs after facility operations end. Regardless of a radiation organization's specific structure, in order to control exposure effectively its program should include several features. Overall, this means that the authority and responsibility for radiation protection should be allocated to the highest management level, then emphasized at all supervisory levels in proportion to the amount of radiation exposure expected. The success of a program depends primarily on management's clear commitment to radiation safety. This level of commitment is expressed when management provides adequate human and financial resources to implement programs successfully, instills in employees an awareness of their own responsibility for safety, and evaluates program effectiveness on an ongoing basis. In short, this means that management must apply the same sound management principles to the task of producing an effective safety program that it applies to producing ore or any other end product. In addition, program success depends on making workers aware of the role they play in reducing unnecessary radiation exposures. Specifically, they need to accept the importance of complying with radiation safety rules and reporting potential problems, such as malfunctioning equipment and procedural violations. Tools a r e available for building this compliance: policy statements, training programs and less formal communications, including the example set by supervisors and managers. The training of individual workers to be keenly aware of their own responsibility for radiation safety is crucial to ensuring the effectiveness of any radiation protection program.

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2. DESIGN OF RADIATION PROTECTION PROGRAMS

2.3 Radiation Safety Officer Employing a Radiation Safety Officer (RSO) is warranted for those (few)mineral extraction facilities where exposure pathways are complex and difficult to control. At those facilities, the RSO should supervise the radiation safety program, providing technical advice as needed. To be fully effective, the RSO should report to senior management and have the authority to enforce radiation safety regulations and administrative policies at all levels of the organization. In addition, the RSO should be provided with adequate resources and not be assigned duties which may lead to a conflict of interest where radiation safety is concerned. This level of authority does not, however, imply total responsibility. A radiation safety program is most effective only when everyone involved in operating the facility is committed to the same objective of reducing risk to a level which is ALARA. The RSO's responsibilities, therefore, also include guiding the operations groups so that they consider measuring, evaluating and controlling radiological conditions whenever they are performing either their ongoing activities or planned changes. To be effective,the RSO may need to develop safety rules that are specific to that facility or organization. The RSO should possess a combination of education, radiation protection experience and appropriate training consistent with the magnitude of potential radiation risk and the complexities of the specific program. In some facilities, one person may appropriately perform all the RSO and industrial hygiene or safety functions. Other facilities, particularly those which must deal with more varied sources of radiation, may require a t least one professional with more specialized education and experience.

2.4 Radiation Safety Committee

In some cases, management should establish a radiation safety committee to help define program scope and enhance their ability to review a program's effectiveness. This committee may be set up for any facility but is especially useful in those cases where potential radiation exposures approach the dose limits. The RSO may work with and be an ex-officio member of this committee, normally composed of several people aware of the facility's radiation protection program and needs. For example, the committee might include supervisors of maintenance, production and engineering. It should review radiation safety analyses of operations and facility operating

2.5 PREPARATION AND MAINTENANCE OF RECORDS

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procedures, assessing the need for increased management attention to the radiation safety program. It may also review specific radiation safety issues, providing advice on how to reduce radiation exposure, and work on improving communications between employees and facility management.

2.5 Preparation and Maintenance of Records Systematic record keeping documents the extent to which the radiation safety program has been implemented and provides a means for determining how effective it has been in meeting its objectives. Specifically, this documentation demonstrates that potential exposures have been evaluated and appropriate controls instituted. Its value becomes apparent when questions are raised about how well workers and the public have been protected from unwarranted exposure and the assumed risks associated with that exposure. To be effective, this record keeping should include certain basic data. Records should describe both the radiological conditions found a t the facility and the radiation doses received by workers. In addition, sufficient data should be maintained so that the environmental impacts of the facility can be assessed. It should also be possible to define patterns of radiation levels and exposures for the various modes of the facility's operation. In setting up this records system, management should consider the potential need for interpreting and comparing data among similar types of facilities and against established radiation protection standards and guidelines. The extent of the record keeping system and the types of records maintained vary with the complexity of the radiation safety program, the more complex the program (due to factors such as higher potential exposures, multiple pathways of receiving exposureor multiple radiation sources),the larger the records system required. The types of records which may be generated and retained include the following: (1) information on radiological conditions on the facility site, a. radiation surveys, b. surface contamination surveys, c. airborne radionuclideconcentrations,usually for radon and1 or its progeny and for airborne particulate matter, d. radioactive materials inventory and disposal; (2) evaluations of radiation exposure of workers and visitors, a. effective dose from external radiation and how it was determined,

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2. DESIGN OF RADIATION PROTECTION PROGRAMS

b. committed effective dose from intake of radioactive materials (eg.,by inhalation) and how it was determined; in many cases, the maintenance of individual records of duration of exposure multiplied by corresponding concentration of airborne radioactive material is appropriate, c. bioassay data needed to estimate any uptake of radioactive materials by personnel; (3) evaluations of radiological impact on the environment, a. radioactive emuents to the environment, b. environmental modeling and/or monitoring; this may include descriptions of meteorological, climatological and hydrological data used in modeling and assessment efforts, c. estimates of individual and collective doses to the public; (4) program implementation documentation, a. safety assessments of designs and operations; this may include rationale regarding why extensive radiation control measures were not necessary for the facility, b. descriptions of unusual operational events involving the potential for radiation exposure; this includes descriptions of corrective actions and/or measures taken to prevent recurrence, c. standard operating procedures and relevant corporate policies, d. training course descriptions and rosters, e. quality assurance data; for example, records on radiation measuring instruments and their calibration. Records should be dated to enable reconstruction of the radiation safety program for any time period. Recommendations on the form, content and retention of radiation program records are provided in Report No. 114 (NCRP, 1992). The length of time for retaining these records varies with the document. Records of exposure of individuals may need to be maintained for a t least the lifetime of the individuals. When facility operations are to be terminated, the need for continued retention of records should be evaluated. Three criteria applicable to the evaluation should be applied: (1) Will the records be needed for medical or legal reasons to establish radiation exposure history for individuals? (2) Will the records be needed for legal or administrative reasons to establish the radiological conditions of the site? and (3) Will t h e records be needed to document compliance with regulations? If these questions cannot be definitively answered, the prudent choice may be to retain records until the uncertainties are resolved.

2.6 QUALITY ASSURANCE PROGRAM

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Facility management should consider compiling and publishing the pertinent results of research, modeling and monitoring. This assists the scientific community in its ongoing evaluations and communications of exposure and risk.

2.6 Quality Assurance Program

A quality assurance program for radiation safety is designed to provide confidence among managers, workers and regulators that the radiation safety program is meeting its objectives. The quality assurance scope is broad, encompassing all the activities associated with defining job scope, measuring job performance, and verifying and documenting successful work completion. This means that high quality programs demand high levels of commitment to quality in every aspect of a program: facility design, operating plans and procedures, adherence to those plans and procedures, and verification and documentation of that adherence. When not only qualitative but also quantitative measurements are required, precision and accuracy become prime objectives. Replicating and performing controlled tests of survey and measurement techniques enhance the validity and credibility of results. Calibrations should be performed using sources traceable to the National Institute of Standards and Technology (formerly the National Bureau of Standards) or other recognized standards organizations. In addition, radioanalytical laboratories should participate in a recognized inter-laboratory cross-check program (NCRP, 1991a). Management should also make sure that any ongoing radiation safety program is reviewed and audited periodically. Results of these checks allow management to evaluate program effectiveness, define improvementsthat will better control exposures, and track and document the ways these improvements are implemented. This close surveillance should include the use of frequent inspections. During these inspections, the facility's staff can observe operating practices and review the need for corrective actions. Also, more formal reviews should be conducted periodically, critically assessing pertinent data from surveys and inspections, personnel exposure and training. To get a clearer picture of the program's effectiveness relative to other programs, data should be drawn not only from the facility itself, but also from similar facilities. During these reviews, special attention should be given to identifying temporal trends in results and equipment or procedural problems. These insights can then be used to guide the development of appropriate program changes.

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2. DESIGN OF RADIATION PROTECTION PROGRAMS

In addition to conducting a variety of reviews, management should consider including reviewers with differing training and experience. Program effectiveness and record keeping should be assessed not only by those directly responsible for the program but by outside observers as well. The frequency will depend on the complexity of the radiation safety program. Once the reviews are conducted, results should be communicated to management at levels high enough in the organization to make sure that appropriate follow-through will take place, especially changes in operating practices, staffing levels, training and commitment to radiation safety or resolution of other program deficiencies which have been identified.

2.7 Coordination Among Safety Programs

A radiation safety program should be carefully coordinated with the facility's overall safety program. After all, the programs share the same purpose: To control risk so people will not suffer adverse health effects or lose the ability to do their jobs. To accomplish that purpose, radiation safety professionals use methods similar to those of other health and safety professionals. This means that, when appropriate to the types and levels of risk, the same people may perform both radiation and other safety activities. In any case, the radiation safety program should complement the other health and safety programs. Content for a specific program can be developed using four basic approaches: (1) consolidating data on risk; (2) developing a perspective by comparing, for example, radiation risks to other risks; (3) defining an orderly process for assessing risk-benefit relationships; and (4) creating a means to perform those tasks effectively (NCRP, 1980a). In some cases, defining an appropriate control program should include efforts to evaluate separate components of risk simultaneously. For example, in evaluating the control needed for airborne uranium ore dust, exposure to uranium and silica must both be considered. Similarly, in evaluating the risk associated with splashing of some solutions, both the pH and the concentrations of radionuclides and their decay products must be addressed. Cases may arise which call for protective measures for radiological purposes that differ from, or even potentially compete with, those warranted to

2.7 COORDINATION AMONG SAFETY PROGRAMS

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mitigate nonradiological risks. In those rare cases, maximal protection against overall risk should be provided. Radiation and other health and safety professionals should work together, assessing the combined risks along with costs of control, then determining appropriate action (IAEA, 1987). Cooperation is imperative in two areas of risk control: demonstrating management commitment to safety and instilling in employees an awareness of their own responsibility for safety. Just as development of a total radiation safety program depends on assessing the impact of specific conditions, medical surveillance, based on the general principles of occupational medicine, should take into account specificworking conditionsand the potential radiation and other risk factors at the particular work site (ICRP, 1986).

3. Sources of Potential Radiation Exposures The design and implementation of radiation protection controls in mineral extraction are influenced by a variety of factors,including ore1 type, ore distribution and quality, mining method, extraction process, land use characteristics, geology and hydrology, and the radiation situation. The discussion of the natural radiation environment in this Section brings forth an awareness of the potential for radiation exposure for any extraction operation or process. The following discussions of potential pathways to individual exposure (Sections 3.2, 3.3 and 3.4) help relate the existence of naturally radioactive materials in the workplace to those locations and operations for which radiation control practices may be appropriate.

3.1 Source Characterization 3.1.1 Naturally Occurring Radioactive Materials

Naturally occurring radioactive materials of concern in the mineral extraction industry are predominantly associated with the 238U (uranium) and 232Th(thorium)radioactive decay series. These radioactive decay series are described in Appendix A. A thorough description of natural radioactivity is found in NCRP Report No. 94 (NCRP, 1988a). From the contents of that report, it can be seen that the abundance of uranium and thorium varies widely over geographic areas. Igneous and sedimentary rocks on average contain concentrations on the order of 0.5 to 5 mg per kg (conventionally)of 238Uand 2 to 20 mg per kg of 232Th.These correspond to radionuclide concentrations of about 6 to 60 Bq per kg of 'The word "ore"in this Report may refer to either a natural combination of minerals from which an extraction is to occur or a technologically altered combination of minerals from which additional extraction is to occur. An example of the latter type of operation may be the extraction of tin from a slag residue resulting from a previous processing operation.

3.1 SOURCECHARACTERIZATION

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238Uand 8 to 80 Bq per kg of 232Th.In the absence of chemical or physical separation processes, an equilibrium is reached in which the number of atoms of each nuclide of a radioactive series that decays during a specific time interval nearly equals the number of decays of the parent nuclide in the series. The activity of each member of the uranium series, for example, would therefore be about 6 to 60 Bq per kg in rock. Chemical and physical separation are common, however, primarily due to mechanical processes and the effects of water movement through the rock. These processes lead to the relative depletion of some nuclides from certain rocks and soils and the relative concentration of others. While such concentration and depletion processes create the potential for economical extraction of some minerals, they also result in widely varying radionuclide concentrations over ore types and locations. In the mining and milling industry, personnel extract a mineral which is relatively concentrated.In the process, they may contact and receive radiation exposures from nuclides of the naturally occurring radioactive series. The magnitude of exposure depends on the characteristics of the specific mineral strata and the extraction processes. Examples of naturally occurring radioactive materials in mineral resources are listed in Table 3.1 (Gesell and Prichard, 1975; Gesell et al., 1977; CRCPD, 1981; NCRP, 1988b; Drummond et al., 1990; IAEA, 1990; EPA, 1991; Johnston, 1991; Pinnock, 1991). The list should not be considered all-inclusive. For example, exposures have been attributed also to smelters processing lead and zinc (NCRP, 1987a) and may occur in the extraction of virtually any mineral. In mines, sources of radiation exposures may include external gamma radiation and airborne radon, radon-decay products (progeny) and ore dust containing radionuclides. Radon and its shortlived progeny often constitute the most important potential exposure source, but the long-lived alpha-emitting materials in ore dust are also of concern. The ingestion of radioactive materials and exposure to external gamma radiation warrant consideration but generally are of less significance (IAEA, 1976a; ICRP, 1977; 1981). Sources of exposure during the extractive processes are similar to those in the mine environment. However, milling can result in elevated concentrations of various radionuclides at different stages of the process. Consequently, the potential for exposure to airborne radionuclides and to external gamma radiation can be increased relative to that experienced in the mines (IAEA, 1976a). The uranium and thorium series are also associated with petroleum and natural gas deposits. Consequently, 222Rnmay be present in natural gas, and 'lOPb and 210Pomay be present in condensed

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TABLE3.1-Natumlly occumng mdioaetivity related to mineml resources. Mineral Mineral or waste radioactivity Aluminum 0.25 Bq Ulg ore 0.1-0.4 Bq Ralg (bauxitic limestone, soil) 0.03-0.13 Bq W g (bauxitic limestone, soil) 0.7-1 Bq Ralg (tailings) 0.03-100 + Bq U/g ore 0.02-0.11 Bq W g ore Fluorspar Uranium series 4 Bq Ralg (tailings) Iron Uranium series Thorium series Uranium series (tailings) Molybdenum Monazite 6-20 Bq U/g sands Thorium series (4% by weight) Natural gas 2-17,000 Bq Rn/m3 (gas, average for groups of U.S. and Canadian wells) 0.4-54,000 Bq Rdm3 (gas, individual U.S. and Canadian wells) 0.1-50 Bq 21?Pb,2'0Po/g(scale, residue in pumps, vessels, and residual gas pipelines) Uranium series Niobium (columbium)Thorium series tantalum Oil Bq RaIL, ranging from mBq to 100 BqIL (brines or produced water) Bq Ralg, ranging up to 70 Bqlg (sludges) Bq to tens of Bq Ralg, ranging up to 4,000 Bqlg (scales) 100-4,000 mBq U naturallg ore Phosphate 15-150 mBq Th naturallg ore 0.6-3 Bq Ra/g ore Thorium series Potash Potassium-40 Uranium series Rare earths Thorium series Tin 1-2 Bq Ra/g (ore and slag) 30-750 mBq Ulg ore Titanium 35-750 mBq Thlg ore 15 Bq Ra/g (ore) Uranium 100 Bq Ralg (slimes) 10-20 Bq Ralg (tailings) Vanadium Uranium series Uranium series Zinc Thorium series 4 Bq U/g sands Zirconium 0.6 Bq Thlg sands 4-7 Bq Ralg sands

3.1 SOURCE CHARACTERIZATION

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liquids in residues in pumps, vessels and residual gas piping associated with natural gas processing plants. In addition, various members of the series, principally "=Ra and 228Ra,accompany produced fluids (production water) and entrained sand to the surface where they may be present in the production water, in scales formed in piping and equipment and in sediments found in tanks, equipment and ponds. Under some circumstances there may be the potential for exposure of operating personnel to external gamma radiation andlor airborne radon, and of maintenance and support industry service personnel to external gamma radiation, airborne radon andlor 210Pb/210P~ surface and airborne contamination. The radioactive materials also present challenges for the appropriate disposal of fluids, scales and sediments. Still another consideration is the potential for exposure to segments of the general population through salvage and reuse of contaminated piping and equipment. The term "elevated is used in this Report to describe naturally occurring radioactivity that is several times greater than average background. The term "enhanced" means the increase over background that occurs when an area is disturbed by human activities. No new radioactivity is produced, but the radionuclides are redistributed in a way that increases the actual or potential human exposure. At present, exposures due to naturally occurring radioactivity are not subject to uniform regulation, but the recommendations of this Report apply to all such exposures in the mineral extraction industry (NCRP, 1984a). Radionuclides released to the environment through air and water can result in radiation exposures to the public. In some cases, final products, process by-products and wastes containing radionuclides have been released for unrestricted uses resulting in unnecessary radiation exposures (NRC, 1980). 3.1.2 Distribution of Radioactivity in Ore, Product, By-Products and Wastes

Managers of mineral extraction industries should determine the distribution of radioactive material at the various stages in the mining, milling and beneficiation processes. The extractive process can often significantly alter the distribution of radioactivity from that in the original ore, thus having a significant effect on the potential for, and type of, radiation exposure at different locations within a plant. Identification of the concentrations of radioactive materials in processes, in the final products or process by-products and in

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waste streams is important for radiation protection and responsible decision making. In addition to bringing the radioactive material physically closer to humans, ore processing may change the chemical availability of the radionuclides with respect to water solubility, plant uptake and metabolic behavior. The nature of most chemical processes is such that solubility and similar factors are increased (NCRP, 1984b); however, solubility changes, for example, vary with the extraction process and the radionuclide. Examples of materials containing radioactivity that are generated by extractive processes are animal feed supplements, fertilizers, dental polishing material, soil conditioners, construction fill, wallboard and foundation base, and mine and mill wastes including some slimes. The concentrations of radioactive material in these materials vary widely, such as 0.4 to 2 Bq 226Raper g of solids, 0.03 to 40 Bq '"Ra per L of liquid, and 2,000 to 4,000 Bq 238Uper L of liquid (Gesell and Prichard, 1975; CRCPD, 1981; Dixon and Hipkin, 1983). Recently compiled data on construction materials and mining and agricultural products appear in NCRP Report No. 95 (NCRP, 1988b) and indicate values from a few to a few thousand mBq 238Uper g of material. Similar levels of 230Th,232Th,226Raand 'loPo are reported. Typical soil in the United States, for comparison, contains about 40 mBq each of '"Ra, 238Uand 232Thper g, and the U.S. Environmental Protection Agency's (EPA) surface soil standard for remedial actions a t inactive uranium processing sites is 185 mBq n6Ra per g (EPA, 1983a; NCRP, 1988a; 19884.

3.1.3 Characteristics Related to Radiation Dose The potential radiation dose that workers or the public may receive from exposure to radioactive materials is determined by a number of factors. These include the amount of the material involved, the types of radiation emitted by the material, the chemical and physical form of the material, the solubility of the material, the particle size distribution of the material, the duration of the exposure, the amount of material that may be resuspended from past releases, the dispersion and dilution conditions a t the time of exposure, the ingestion pathways involving contaminated water, food stuffs and animal feeds, and the demographic and physiological characteristics of the population exposed (ICRP, 1979-1988; 1982; 1991; NCRP, 198413). 3.2 Occupational Exposures 3.2.1 External Radiation External sources of radiation exposure to mine workers are caused by the concentrations of naturally occurring radioactive materials

3.2 OCCUPATIONAL EXPOSURES

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present in the deposited mineral, especially the concentrations of the lead and bismuth nuclides that are intermediate in the radioactive series. External radiation levels i n most mines a r e less t h a n 0.01 mGy per h and may not be of concern in specific facilities. Levels can exceed 0.2 mGy per h and approach 1mGy per h in uranium mines in areas of selective deposition or exceptionally rich ore bodies (IAEA, 1976a; ICRP, 1977). High-concentration uranium- or thorium-bearing ores will have gamma and beta radiation fields present during crushing, grinding and stockpiling operations prior to any physical or chemical processing. During the processing of ores, chemical reactions can result in changes in the relative concentrations of the naturally occurring radioactivedecay products that are present. The concentration of specific radionuclides can be increased; this enhanced concentration depends on the nature of the process, the chemical and physical characteristics of the materials, and the associated carrier materials (Gesell and Prichard, 1975; IAEA, 1978). Piping and process vessels, product bins and waste repositories may have concentrations of natural radioactivity sufficiently high to require protection of workers from external radiation exposures. In addition to the naturally occurring radioactivity in the ores, other potential occupational exposure sources can include gauges and other measuring devices that contain radioactive materials or sources of ionizing radiation. These devices are commonly used in flow streams and other areas as part of the process control.

3.2.2 Airborne Radioactivity Airborne radioactivity may occur in both mining and milling procedures. The primary airborne radionuclides in uranium and many other mines are '"Rn and its short-lived progeny, 'laPo (RaA), '14Pb (RaB), '14Bi (RaC) and '14Po (RaC'). In some situations 220Rn(thoron) and its short-lived progeny can be present due to the presence of natural thorium in the ore. The concentrations of radon gas and its progeny can vary widely within the same mineral industry as well as within a single facility. The relative concentration of different radionuclides may also vary with a specific process (ICRP, 1977; 1981; NCRP, 1984a; 1984~). During the milling and extraction processes there may be many situations that provide exposure to airborne radioactivity. These include grinding and crushing operations, ore storage, dry feed transfer, sampling and analytical procedures, plant maintenance, product

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drying and packaging, by-product preparation, waste disposal, transportation and accidents (IAEA, 1976b; 1981). Dust particles may contain 238U,234U,235U,228Th,230Th,232Th, 226Ra,224Raand 210Po.These materials are important because they are alpha emitters which can irradiate internal organs of the body after inhalation or ingestion of the particles. Other radioactive decay products, such as 'lOPb, can also irradiate internal organs of the body as a result of beta and gamma emissions, and as precursors of other radioactive decay products.

3.2.3

Surface Contamination

Transferable surface contamination can be a source of inhalation and ingestion of radioactive materials. Therefore, control of surface contamination is an essential component of any radiological protection program. Operations that may result in contaminated surfaces include crushing and grinding, process equipment maintenance, agitation of solutions, decontamination for the release of equipment for unrestricted use and decontamination in emergency response operations. Of particular importance for contamination surveys are areas such as lunch rooms, storage of products andlor by-products and shipment facilities in which both inhalation and ingestion of radioactive material could occur.

3.3 Releases to the Environment The following paragraphs describe the release points and release mechanisms applicable to mineral extraction facilities. Subsequent sections of this Report present additional information on effluents and the environment. A figuredescribingprincipal radiological exposure pathways to humans from mineral extraction operations is contained in Section 6 (Figure 6.1) and is useful in perceiving the relationships between releases to the environment and the radiation exposures of members of the public. 3.3.1 Airborne Radioactive materials may be released to the ambient atmosphere through a number of mechanisms during mining, extraction processes, final product and by-product preparation, waste disposal, effluent discharge and transportation.

3.3 RELEASES TO THE ENVIRONMENT

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Mine ventilation discharges may contain radioactive material, particularly 222Rnand its short-lived radioactive decay products. Particulate concentrations, including the long-lived radionuclides of the radioactive uranium and thorium decay series, also are present. Dusts having elevated radioactive material concentrations may be resuspended from mine haulage roads. Mill exhaust stacks may release significant amounts and concentrations of the radioactive material identified in Section 3.2.2, depending on the operations. These operations may include crushing and grinding; sampling and analysis; transfer and conveyance; leach, extraction or separation functions; product and by-product preparation, drying and packaging; and waste disposal or retention. High-temperature processes such as calcining or sintering may result in releases of radioactive materials to the atmosphere. Ore stockpiles can also be a source of airborne emissions (NRC, 1980; NCRP, 198413; 1984~). Incidents involving spills or other releases during transportation of raw materials, finished products or by-products can introduce radioactive materials to the environment and expose the general public.

3.3.2 Waterborne

The releases of radioactive effluents in sufficient amounts and concentrations to water bodies, waterways and subsurface aquifers can result in elevated radiation exposures to impacted populations. These waterborne radioactive releases can occur through planned discharges to the environment, from improperly designed or managed impoundments, or during accidents. Reactions can occur between process waste liquids and an impoundment's natural or membrane liners, resulting in degradation of the ability of the liner to contain the liquid. Liner degradation can also occur from interactions between leachate and underlying soils. Such degradation depends on the nature of the radioactive materials involved and the potential chemical and physical interactions of the waste liquid. Proper management of impoundment discharge pipes, beaches and freeboard are also important factors in preventing unintentional releases (IAEA, 1981). Examples of sources of potential waterborne discharges are mine dewatering, aqueous effluents from ore crushing and sorting efforts, mill waste water processing, raffinate and tailing transport and disposal, process vessel and piping rupture, and retention system failures (IAEA, 197613; NRC, 1980; EPA, 198313).

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3.3.3 External Radiation

Releases to the environment can increase external radiation levels. These levels could be from ground surface contamination as a result of prolonged airborne releases or impoundment failures that produce discharges of radioactive materials to surface and ground waters. External radiation exposure may also result fiom "shine" (radiation scatter) that can occur from large sources such as a tailings pile or ore stock piles, and from transportation incidents.

3.4 Process By-Products and Waste Materials Uranium and thorium ore mining and milling generate wastes which have most (-85 percent) of the original radioactivity of the ore. Other industries (such as phosphate, copper, fluorspar, vanadium, bauxite, titanium and rare earths operations) process ores which often contain elevated concentrations of uranium, thorium and their radioactive decay products. Examples of such materials are provided in Section 3.1.2. Radiation exposures may result from the release, for unrestricted use, of a radioactive process by-product, radioactive waste materials, or contaminated equipment, materials and land. Due to the possible wide use of such materials, the resultant collective radiation dose for a segment of a population may become significant. Examples of unrestricted use situations in the past are the use of uranium mill tailings in construction, the use of elemental phosphorus plant slag in roads and building material, building construction over phosphate mined lands, fabrication of wallboard from gypsum containing elevated levels of radioactivity, and the manufacture of tape dispensers using zircon sands as ballasts. These examples illustrate a wide range of practices involving potential radiation exposure. The scope of this Report does not include evaluations of the various possibilities for material recycling. Should facility management choose to explore the reuse of materials, careful evaluations must be made to determine the radiation exposure potential from the release of any product, process by-product or waste materials. Once the exposure potential has been defined, decisions can be made for restrictions or controls on use to ensure that exposures are ALARA. Coordination with the appropriate regulatory authority is advisable due to the changing nature of regulatory practice (Gesell and Prichard, 1975; CRCPD, 1981; NCRP, 1988b).

4. Exposure Management Program The objectives of the radiation exposure management program are to ensure that individual workers and members of the public do not receive radiation exposures resulting in doses that exceed recommended limits and to maintain all radiation exposures a t levels as low as reasonably achievable. Principles and methods important to these objectives are discussed in this Section and include the basic program elements of design and engineering features, operational practices and training. The principles and methods discussed below include elements which are optimally a part of the management decision making often thought to be applied for purposes other than radiation protection. To improve the cost-effectiveness of operations, facility management often develops design and operational features to optimize the equipment types and numbers of worker hours required to produce the end product safely and reliably. Proper facility and work planning contribute to radiation exposure management when they lead to the lowest reasonable number of worker hours near radiation sources.

4.1 Exposure Limits

The NCRP recommends that annual radiation exposures of workers do not result in a n effective dose exceeding 50 mSv and do not approach that value unless the implementation of further exposure control measures is clearly not reasonable. NCRP's guidance for cumulative exposure of workers is that the effective dose should not exceed 10 mSv times the individual's age in years. For members of the public, the recommended annual limit is 1mSv where exposure is continuous or frequent and 5 mSv where exposure is infrequent (NCRP, 1993).I t is important to recognize that the above limits are based on the sum of internal and external exposures. These limits are of a boundary nature and, for routine operations, are the upper limits that should be approached only when the cost of further exposure reduction is unreasonable. This means that even though exposure conditions for most segments of the minerals

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EXPOSURE MANAGEMENT PROGRAM

extraction industry are unlikely to cause any individual's exposure

to approach the upper limit, optimizing the protection of workers and the public, in accordance with the principle of ALARA, is appropriate throughout the industry. Simply stated, it is incumbent upon management to assure that all exposures are ALARA, economic and social factors being taken into account (NCRP, 1993). Additional information on guidelines, standards and potentially applicable federal and state regulations is presented in Section 7.

4.2

Exposure Environment

Mining and milling activities pose a potential for exposure to external gamma radiation and, to a lesser extent, beta radiation and to airborne radon, radondecay products (progeny) and larger particles (eg., ore dust). The exposure potential varies significantly across the different minerals extraction operations and depends upon factors such as geological formations;type, distribution and quantity of ore; and mining and processing methods. Exposure potential typically is greater in underground uranium mining activities than in any of the other minerals extraction operations. Typical radiation environments for underground uranium mines and associated milling are described below, because they exemplify the more complex radiation control requirements. 4.2.1

External Radiation

External exposure results from radiation emitted during the decay of uranium and uranium progeny. The exposure rate varies with location in the mine and the milling operations, being higher in the ore bodies than in barren rock and relatively high in process areas where radium is concentrated. In mines, the gamma exposure rate per h in shafts and tunnels to 5 to varies from less than 1 F G ~ 15 pGy per h near ore faces for ores of 0.2 percent uranium oxide.2 Very high grade ores, on the order of 20 to 30 percent uranium oxide, have gamma exposure rates on the order of 1mGy per h or greater (ICRP, 1977; 1986). Beta exposure fields are of less importance qn this Report, radiation intensity (exposure or exposure rate) is presented in terms of air kerma (see Glossary), expressed in units of Gy or Gy per h (rad or rad per h or R or R per h in the conventional system of units, see Appendix B). At this time, instrument value may be expressed in any of these units. 1 Gy = 100 rad 100 R.

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although beta radiation dose in air near the ore surfaces may be higher than the gamma dose by a factor of ten (FRC, 1967). The wide distribution of radiation sources in mines and high gamma energies associated with the decay series generally result in uniform whole-body exposures. The ratio of gamma to beta exposures in combination with the lower limit on whole-body exposure versus exposure of the skin (or extremities or lens of the eye) generally means that gamma irradiation is more important than beta irradiation in defining exposure control measures in mines (ICRP, 1986). In uranium mills, higher radiation levels (on the order of tens of kGy per h for typical ores) may occur in the vicinity of yellowcake drying and packaging areas, above some of the ore storage piles, and in the vicinity of tailings. Sources of gamma radiation may also exist from accumulations of 226Raand its radioactive decay products in certain piping and process vessels. Consideration should be given to the potential for higher radiation levels associated with residues in plant equipment in uranium mills and other plants involving chemical processing of minerals. The relative concentrations of the naturally occurring decay products change during ore processing. Concentrations of specific nuclides may be relatively depleted or enhanced a t specific points in the process stream. However, ingrowth of the beta and gamma emitters will occur in a relatively short period of time in storage areas for finished products such as yellowcake (IAEA, 1976a). 4.2.2

Ore Dust

Ore dust containing uranium and uraniumdecay products may be suspended in the mine air due to blasting, ventilation and other mining operations. Suspension of ore dust also may occur during ore transport, in scale houses and during ore crushing. The concentrations of the radionuclides are quite variable with time, location and the activities underway. In ordinary mining activities, the average concentration of the radionuclides is below 0.3 Bq per cubic m of air (ICRP, 1977) and therefore, on average, below the limiting airborne concentrations specified in regulations (Section 7).

4.2.3

Airborne Radon and Radon Progeny

The most significant source of radiation exposure in the uranium mine environment is radon, a noble gas that is produced by the decay of radium in the ore bodies. Concentrations of 222Rnin outdoor air

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4. EXPOSURE MANAGEMENT PROGRAM

average approximately 0.008 Bq per L (NCRP, 1988~). In mines, the concentration of radon can range to high levels-on the order of tens to hundreds of Bq per L-depending upon the rate of emanation from the ore, the mining activity and the ventilation. The decay of radon can result in the buildup of high concentrations of radon progeny, on the order of tens to hundreds of working levels (WLs) (see Glossary) depending again upon ventilation and mining activities. The inhalation of radon progeny represents the principal radiation hazard and results in internal exposure to bronchial tissues (NCRP, 1984~). In mills and other extraction facilities, radon and its short-lived progeny may also build up to concentrations requiring the application of exposure control methods. Higher concentrations are found in ore storage bins and other areas with lower air turnover rates.

4.3 Facility Design and Engineering Other than the magnitude of the levels, the characteristics of the radiation environment described above for underground uranium mines are similar in other minerals extraction operations. Management has the responsibility for determining the need for exposure management practices in any particular operation and for developing and implementing programs to provide adequate radiation safety. The most effective approach in meeting exposure control objectives is through proper design and engineering of facilities. Such facilities allow for a much higher degree of safety than can be obtained by dependence upon administrative rules and procedures. In addition, the operating difficulties that might be imposed as a result of radiation exposure or safety requirements can be minimized. Consequently,design and engineering features should receive careful consideration in the planning of new operations and modification of existing facilities. Input at the early planning stages by individuals knowledgeable about radiation protection requirements will ensure that proper radiation safety features are incorporated in areas such as site selection, facility layout and equipment, and system design.

4.3.1 'Site Selection The specific location of facilities should take into account factors such as the geologic, meteorologic and hydrologic characteristics of the area as well as the type and quantity of radioactive materials

4.3 FACILITY DESIGN AND ENGINEERING

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that may be involved and the potential for exposure to and release of those materials. The mining operation obviously has to be located at the ore body. Surface facilities should be located to keep the potential for unnecessary worker exposures and for release of radioactive materials to the environment as low as reasonably achievable. This should be done by locating the ore storage pads away from normal traffic areas to prevent casual and unnecessary contact by workers, and locating the pads in areas protected from surface winds and runoff waters to reduce releases to the environment. Release ducts for mine ventilation shafts should be installed vertically to obtain maximum dispersion and dilution of exhaust air and should be located downwind of the surface facilities and away from populated areas. Ore processing facilities ideally should be located close to the source of the ore. This arrangement reduces the distance over which the ore has to be hauled, thereby reducing the potential for loss of control (e.g.,spills) and release to the environment. In addition, mine drainage water, when available, can be used in the mill process and consolidated with the mill liquid effluent. Consideration should also be given to the proximity of the off-site population, prevailing wind patterns and site geologic and hydrologic characteristics. 4.3.2 Facility Layout

The proper layout of facilities is a critical factor in reducing the likelihood and magnitude of exposure to radiation. If the material being processed has the potential for causing exposures approaching the nonoccupational exposure limits, the facility should be designed to isolate areas where the exposure potential is increased, to maintain separation of contaminated and noncontaminated areas and to control the movement of personnel into and out of those areas. These control measures may also be appropriate as the concept of ALARA is incorporated into the facility design process. The layout of an underground mine is dictated by the ore pattern and the mining method employed. Nevertheless, the design should provide for establishing clean areas for lunchrooms and maintenance shops, isolating inactive or mined out areas, and ensuring adequate ventilation in all working areas and passageways. The layout of a surface mine should take advantage of site meteorologic and topographic features so that mining activities are predominantly downwind of the support facilities and precipitation runoff is channeled around or away from the mining activities and any populated areas not directly associated with the mining activity.

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The layout of processing facilities normally provides greater opportunity for incorporating exposure reduction features than does the design of the mine. Design features that should be considered include the isolation of process vessels and activities where radionuclides would be expected to concentrate, such as the uranium drying and packaging circuit in a uranium mill. The layout should provide adequate separation of process areas from nonprocess areas and should group dust-producing activities, such as ore crushing, in the predominant downwind direction from other activities. 4.3.3 Equipment and System Design

Specific equipment and system design features can significantly reduce the likelihood and magnitude of exposure of individuals to airborne and external sourcesof radiation, and the release of radioactive materials to the environment from operations and wastes. The specific combination of equipment and systems features for exposurereduction purposes will vary greatly across the minerals extraction industry and will generally be more comprehensive for uranium mining and milling operations. As noted in Section 4.2.3, the principal airborne exposure concern for most underground mining operations is radon and radon progeny. Mechanical ventilation is the most effective manner for controlling exposure to these radionuclides. Radon enters mine air along ventilation passageways and remains entrained until the ventilation system discharges the air above ground. The concentration of radon and its short-lived progeny thereby increases continuously with time as air travels through the mine. Optimizing the residence time of air within the mine is a key factor, necessitating that provision of adequate ventilation equipment and systems be taken into account from the initial design and development stages. Ventilation systems should consist of a primary system with high-capacity fans for moving air through main tunnels connected with vent holes and a secondary system that delivers air from the primary airways to active working areas. The secondary system should be modified as required, using small auxiliary fans and flexible ducts, to ensure proper air distribution where individuals are present. Overall ventilation system efficiency may be enhanced by sealing off inactive and abandoned mine areas and by using doors and bulkheads to channel air through designated areas. The use of air cleaning systems to recondition air for further use underground may be evaluated as an exposure control measure (ICRP, 1986). For instances in which inactive mine areas cannot be effectively isolated from active areas,

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additional ventilation systems or altered configurations may be evaluated to ensure adequacy of exposure control in the active areas. To minimize the in-growth of radon-decay products, the mine ventilation system should be designed and operated to ensure the efficient use of air. Recommendations for mine ventilation systems can be found in International Commission on Radiological Protection Publication 47 (ICRP, 1986), International Atomic Energy Agency Safety Series No. 82 (IAEA, 1987) and various mine engineering manuals. Exposure to ore dust may conceivably be the principal airborne exposure concern. This is more likely to occur for surface mines than for the underground mine. Should such a situation arise, the use of enclosures and local ventilation equipment should be evaluated as a means to reduce the airborne radionuclide concentrations near the miners. Also, the use of equipment using water or other sprays for dust suppression should be evaluated. In underground mines with high-grade ores, ventilation flow rates may require optimization, to reduce dose from inhalation of radon-decay products while maintaining acceptable concentrations of resuspended ore dust (Brown,1992). The equipment and system design measures for protecting against airborne radiation sources in mills differ in degree and type from those in mine operations. Airborne radon progeny may build up in locations such as ore storage bins, grinding and crushing areas and at various processing vessels. To reduce the potential for unnecessary exposures, local ventilation equipment should be used to exhaust the air away from working areas and to the effluent treatment system and facility vents. To the extent possible, dry process material should be confined to preclude generation of airborne particulates and if practical, crushing and grinding equipment and conveyor transfer points should be enclosed. Similarly, the final product line and packaging equipment may need to be enclosed and, in the case of yellowcake (uranium) drumming, t h e area should be enclosed and maintained under negative pressure. Since most uranium ores do not present an external radiation exposure hazard, specific equipment and system design measures to protect against external exposure are generally not warranted nor are they feasible for underground mining activities. Only for ore bodies with uranium content above about one percent are special tools or mining methods likely to be reasonable for control of external exposures in mines. However, at various steps in the milling process, radionuclides may be concentrated or accumulated such that increased external exposure potential may be present. Where such a potential exists, the equipment should be designed for accessibility,

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ease of maintenance and ease of installation and removal, and should incorporate other features to reduce the time the worker is required to spend in the vicinity of the equipment. The component should also be constructed, when feasible, of materials that minimize the buildup or collection of process radionuclides. The solid and liquid wastes produced in the minerals extraction industries are potential sources of exposure to workers and of contamination of the environment. Equipment and system design features should be tailored to ensure that these exposure sources are controlled and waste materials are disposed of properly. Groundwater in underground uranium mines is a major source of radon (IAEA, 1987) and may contain other radionuclides that can be a source of environmental contamination upon discharge to the surface. Ventilation measures should adequately control the radon released from mine waters, however, the release underground can be minimized by use of pipes for conveying the water to pumps and pumping stations and enclosing the pumping stations. The radionuclides contained in the mine water can be removed, or their concentrations reduced, through treatment systems such as ion exchange media and chemical precipitation. The balance to be evaluated is minimizing release to the environment versus minimizing exposure of personnel operating and maintaining the equipment in which the radionuclide inventory accumulates. If evaluations indicate treatment systems a r e practicable (consistent with t h e concept of ALARA), equipment and systems should be installed to treat the water. Also care should be taken to ensure that drilling and production-facility shafts are properly designed to minimize the potential for interaquifer exchanges of waters containing radioactive materials. Milling operations produce large quantities of both solid and liquid wastes. The solid wastes are tailings, which contain most of the radionuclides that were present in the ore, and worn out, replaced or obsolete equipment, debris, residues and other materials with radionuclide contamination. The tailings should be handled and distributed with remote equipment, where possible, to minimize handling by individuals. Spraying equipment may be required t o minimize the drying out of tailings and the resultant dust accumulation and to control dispersion of radionuclides. In some cases, covering the wastes to control the spread of contamination may be practicable. Appropriate survey equipment should be provided for determining the extent and type of contamination on and in the other solid waste materials. Depending upon the degree of contamination, isolated storage areas may need to be provided, with access to these areas limited.

4.4 FACILITY PROCEDURES AND PRACTICES

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Mill equipment containing liquids should be covered to prevent generation of airborne aerosols due to splashing. Liquids should be discharged through pipes rather than through open conveyances. Equipment and systems such as ion exchange media, chemical precipitation, and settling and evaporation ponds should be used to treat liquid waste streams containing concentrationsof radionuclides that exceed unrestricted release limits. Use of such equipment and systems also would be appropriate should cost-benefit evaluations indicate that the control measures are reasonable.

4.4

Facility Procedures and Practices

Operational procedures and working practices in both mining and milling activities should minimize potential worker exposures and releases of materials to the environment. Well-defined and written procedures and practices, properly presented to and understood by employees, are an important component of an effective exposure management program. Operational procedures and practices that should be considered in reducing potential exposure include access control, general housekeeping, proper operation of control equipment, management of radioactive materials and use of personal protection equipment. The work practices and procedures should be reviewed and evaluated periodically to incorporate changes that are necessary due to process modifications or changes in regulatory requirements, or to reflect improvements in radiation safety practices.

4.4.1 Access Control One of the simplest and most effective operational practices that can be used to reduce radiation exposure is to limit personnel access to areas where radioactive materials are located. One means of access control is erection of physical barriers preventing access, for example, the construction of walls to keep personnel from unventilated, abandoned work areas. For areas which may be routinely occupied, the use of signs and other "postings" is appropriate, and routine access should be restricted to personnel required for production and maintenance functions (NCRP, 1986). Areas in which dose or airborne concentration limits may be approached should be designated as requiring special precautions for entry; for example, the yellowcake packaging areas at uranium mills. Special work permits should be used to control access to these

32 / 4. EXPOSURE MANAGEMENT PROGRAM areas and to specify additional requirements for personnel performing operations, maintenance and repair work. As a component of operational procedures, visitors to a facility may be required to have an escort. Nonemployee personnel, such as contractors performing work at a facility, should be provided with specific instructions and appropriate training regarding procedures that are to be followed. 4.4.2 Radioactive Material Control

Sound work practices and procedures should be employed to provide positive control over materials containing radionuclides. To reduce unnecessary exposures, requirements should be established for activities such as movement, storage, processing and packaging of materials, waste management, and control of sealed radioactive sources as in gauges and level sensors.

Materials Handling. Almost all stages of ore handling and processing, such as hauling, storing, grinding and crushing, conveying and packaging, afford opportunities for airborne releases and spills. Work practices, specifically good housekeeping, are essential to preclude unnecessary exposure to and dispersion of radioactive materials. All spilled materials should be promptly cleaned up and water or other dust-control measures (for example, curtains or vacuum systems)should be used wherever dust-creating activities occur. Procedures should require that covers be in place for vessels requiring covers, that containers and conveyors not be overfilled to prevent spillage, that filters and dust collectors be maintained regularly, and that inspections be performed periodically to ensure that the prescribed work practices and procedures are being followed.

4.4.2.1

Waste Management. Waste management practices should be designed to move process waste directly from the point of generation to a secure control or disposal location, preferably the final onsite location. Where feasible, waste management may be enhanced by the use of consolidated control or disposal locations versus the use of many such locations. Procedures should include requirements for routine inspections and monitoring to ensure that waste handling systems are operating as designed, that unanticipated exposure of workers is not occurring and that any releases to the environment are within allowable limits and ALARA.

4.4.2.2

Sealed Source Control. Calibration devices, level gauges and other measuring devices containing radioactive materials are commonly used in the minerals extraction industry. Procedures

4.4.2.3

4.4

FACILITY PROCEDURES AND PRACTICES

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should be developed to ensure that such devices are installed, used and maintained only by authorized personnel and in accordance with exposure-control principles, applicable manufacturer's guidelines and regulatory requirements. Labeling, posting, leak testing, operating, storage and disposal procedures should be available for each type of source a t a facility. A radioactive source inventory system should be maintained and the location, condition and use of the sources should be confirmed by periodic inventory audits. 4.4.3

Personnel Protective Equipment

The primary reliance for radiation safety and control should be placed on properly designed facilities and engineered controls rather than on personnel protective equipment. However, there are occasions in which use of protective equipment provides an adjunct to engineered controls. There are also situations in which engineered controls cannot reasonably be provided and the use of such equipment is needed.

Respiratory Protection. Respirators are necessary in some circumstances, such as during a ventilation failure or when performing work in an area for which adequate ventilation is not available. The use of respirators requires procedures for proper fitting and training of those who will wear them and for cleaning, maintaining and inspecting the devices. With the proper selection of the filter medium, respirators can provide effective protection against ore dust, radon daughters and process dust. Depending on the specific work activity, the use of supplied-air respirators and working-time restrictions should be considered a t radon concentrations above about 20 kBq per cubic m (IAEA, 1986). If respirators are used, a respiratory protection program designed in accordance with the American National Standards Institute (ANSI, 1980) should be instituted. Respiratory devices are often uncomfortable and distracting and may impede the wearer's vision. Because of these factors, an individual wearing a respirator may be subject to increased risk of physical injury. Nevertheless, there are situations in which the use of these devices is necessary, such as when performing maintenance on a yellowcake dryer where significant exposures to airborne uranium would be possible.

4.4.3.1

4.4.3.2 Protective Clothing. Protective clothing may be used to prevent contamination of skin and the personal clothing of the workers. An important positive feature associated with the use of

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protective garments is prevention of the spread of contamination. Workers will not inadvertently carry contamination to clean areas if the work practice requires that the garments be left in designated change areas and laundered or disposed of at the facility. The extent of use of protective garments depends upon the potential for contamination. In many cases, only gloves, coveralls and shoe covers are indicated. In other cases, more elaborate measures may be required to protect against contamination.

4.5

Employee Training

All employees whose work involves radiation and radioactive materials should be provided with initial training about the potential for exposure, the risks associated with exposure and the work practices and procedures to be followed to prevent or minimize exposure. The training should include topics such as these from NCRP Report No. 71 (NCRP, 1983): characteristics of ionizing radiation; types of radiation that could be encountered and under what conditions; how exposure might occur, internal and external; basic effects associated with radiation exposure, acute and chronic; basic protective procedures and practices, eg., time, distance, shielding, work practices, protective clothing; radiation monitoring programs, surveys, effective dose, allowable limits; employee and organization responsibilities; and emergency procedures. Written material covering these topics can be valuable for distribution to the employees and testing can serve to document the employee's knowledge of the various topics. Periodic retraining is necessary, usually on a n annual basis and provides an opportunity for employees to ask questions, offer suggestions and express any concerns about the radiation aspects of their work. The extent and breadth of employee training will vary with job requirements and responsibilities. For an individual whose duties do not require presence in the radiation environment, a simple description of the working environment, protective measures to be taken and assurance of personal safety may be sufficient. For those who work, for example, in the final phases of yellowcake processing

4.5 EMPLOYEE TRAINING

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35

and drumming at a uranium mill, a more detailed and extensive training program will be required. The supervisor should be responsible for ensuring that radiation workers receive any additional information required in the form of specific training on the proper procedures for their jobs. The material regarding radiation safety that is discussed in the general training program should be reviewed a t these specific training sessions. The discussion should include the potential radiation hazards, the radiation control devices and the procedures particular to the employee's work, including monitoring techniques.

5. Monitoring of Occupational Exposure As indicated in Section 3, occupational exposure may be associated with external radiation, airborne radon and its short-lived decay products, airborne long-lived radionuclides and surface contamination. Consequently, the design of a monitoring program must consider t h e possible need for monitoring each of these as well a s assessing internally deposited radioactivity in personnel with potential uptakes (bioassay). The exact nature of the monitoring program for any facility will depend upon a number of factors, including type and concentration of the radioactive materials present, type of operations being performed, use and occupancy of the areas being monitored, and actual results of monitoring. This Section addresses considerations generally applicable to the mineral extraction industries, and emphasizes uranium mines and mills as examples of facilities requiring a comprehensive program. Only portions of such a program may be appropriate for other facilities. Other specific types of operations a r e addressed further in Section 9.

5.1 Monitoring Objectives

The general objectives of radiation monitoring are to assess the radiation environment, to evaluate the adequacy of control programs, and to document personnel exposure and dose. Radiation monitoring includes two broad categories, characterization of the workplace and personnel exposure assessment.

5.1.1

Characterization of the Workplace

Programs to characterize the workplace involve some combination of two general measurement modes: (1) survey and short-term sampling, radiation survey, contamination survey and "grab" air sampling; and

(2) area monitoring and long-term sampling, deployment of pas-

sive integrating external radiation monitors, monitoring of airborne radioactivity with passive integrating detectors, continuous air sampling with periodic analysis, use of continuous area monitors for external radiation and use of continuous air monitors. Some specific objectives for various circumstances include: (1) to characterize the radiation environment, identify potential exposures, provide input to the design of a control program and determine whether there is a need for personal monitoring in an unfamiliar or new situation; (2) verify on a routine basis the adequacy of the in-place control program and to detect loss of control; (3) to determine any effect on radiation exposure and identify any need for additional control measures following a change in operating conditions or source magnitude or configuration; (4) to verify the adequacy of control or corrective measures after their installation; (5) to obtain information necessary for establishing working conditions, protective equipment requirements and maximum exposure duration ("stay times") prior to a proposed operation with high exposure potential; (6)to verify safety restrictions and verify return of the source to the shielded configuration when sealed sources are being manipulated; and (7) to estimate personnel exposures and doses in the absence of personal monitoring. 5.1.2

Personnel Exposure Assessment

Objectives of personnel exposure assessment include: (1) ensuring that sufficient information is available to assess the need for additional radiation control measures, (2) documenting individual personnel exposures and estimating doses, and (3) detecting task-related exposures not detected by general monitoring of the workplace. 5.2 Monitoring Program

Initial radiation and airborne radionuclide surveys should be performed to define radiation levels and identify any higher radiation and radioactivity locations. These surveys should be sufficient to

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characterize any temporal variabilities in radiation levels. ThereaRer, a facility or area should be monitored periodically a t a frequency determined by the radiation or radionuclide levels and their variability. In general, more frequent monitoring is needed where levels a r e higher and/or variable; less frequent monitoring is required where levels are low and constant. General guidance is suggested in Table 5.1; specific programs should be developed for individual facilities. In addition, special surveys should be performed: (1) to evaluate the impact of any change in design or operating conditions likely to increase radiation levels or concentrations of airborne radionuclides; and (2) to determine the effectiveness of any equipment or process modifications intended to reduce radiation or airborne radionuclide levels.

5.3 External Radiation

Using the exposure rates stated in Section 4.2.1 and realistic occupancy conditions, the actual annual effective doses to workers from TABLE5.1-Monitoring program. Radiation andlor airborne Monitoring action radionuclide level Perform initial assessment and periodic Sustained or average levels lo% of derived occupational exposure limits."

Institute routine monitoring of the workplace.

Cumulative dose, average concentration, or annual intake to reach or exceed 10% of occupational limits.

Perform individual exposure assessment: Gamma radiation-provide personal monitors. Airborne radionuclide-use personal samplers or assess on the basis of area sampling and exposure time.

Individual measurements or samples approach or exceed derived occupational limits."

Repeat the measurement.

Provide intensive surveillance and monitoring in conjunction with reevaluation of dose-reduction planning. "Derived limit = radiation level or concentration that would result in the annual cumulative limit assuming full-time occupational (2,000 h per y) exposure. Confirmed level approaches or exceeds occupational exposure limits.

5.3 EXTERNALRADIATION

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gamma radiation sources in most mineral extraction facilities have been calculated to be less than 10 mSv per y.3 Higher radiation levels may be expected, however, in association with high-grade uranium ores and in some areas of mills (Section 4). An external radiation monitoring program should provide an overall assessment of radiation levels, identify the locations of the higher radiation levels and provide radiation dose estimates for the individuals in these areas.

5.3.1 Characterization of the Workplace Continuous area monitoring is employed in many radiation facilities. However, in the mineral extraction industries, external radiation monitoring will usually be limited to surveys with portable gamma survey meters andlor integrated measurements with passive gamma radiation monitors. The gamma radiation survey should include each working area in the facility, with particular attention given to fixed working stations or other areas where personnel remain for extended times. Generally, readings should be taken a t waist level at the operator's position in fixed working stations, in the center of general work areas and about a meter from radiation-emitting surfaces. Survey records should clearly indicate these locations. [For a detailed discussion of area survey methods see NCRP Report No. 57 (NCRP, 1978b1.1 Extraction and chemical plants should be surveyed to determine whether there is buildup of radium in residues and sediments in piping and equipment. Thereafter, a survey should be performed prior to maintenance operations or after disassembly, cleaning and repair of any plant equipment containing such radioactive residues. When the personal monitoring program shows unexpected exposures or a trend of increasing levels, surveys should be performed to determine the cause. If doses approach recommended limits, the dominant sources of radiation should be located by meticulous surveys. After corrective action has been taken, additional surveys should be made to verify the effectiveness of the action. Appropriate gamma survey instruments are required. Scintillation survey meters are useful from 50 to 100 nGy per h to several tens of pGy per h. The typical useful range for Geiger-Mueller (GM) survey meters is approximately 200 nGy per h to several hundred pGy per h. Exact ranges for each will depend upon the particular model. Because scintillation and GM instruments can have a dose 31n this Report, effective dose and equivalent dose are expressed in units of Sv and Sv per h (rem and rem per h, conventional system; see Appendix B).

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rate response that is strongly energy dependent, GM tubes should be energy compensated and both types of instruments should be calibrated as nearly a s possible to the energy spectrum of interest. Most ion chamber survey meters are useful a t exposure levels corresponding to 10 pGy per h and above, although some instrument models may measure down to 1pGy per h. A pressurized ion chamber survey meter that measures down to about 50 nGy per h is commercially available. A properly designed ionization chamber instrument has the advantage of having a dose rate response that is relatively independent of energy over a wide energy range. [Detailed information about instrumentation may be found in documents such as NCRP Report No. 57 (NCRP, 197813) and Radiation Detection and Measurement (Knoll, 1989).] Survey instruments should be calibrated before being put into use and after each repair. Operational checks with a radiation source should be performed on each scale that is to be used prior to use of the instrument. Survey meters that are to be used to estimate doses to personnel should be calibrated a t least semiannually. A comprehensive discussion of survey-instrument calibration may be found in NCRP Report No. 112 (NCRP, 1991a). Area radiation also may be monitored by measuring the integrated or accumulated level a t fixed locations, preferably with thermoluminescent dosimeters (TLDs) or alternatively with film badges (see Section 5.3.2). The devices are exchanged and evaluated to estimate dose on a monthly to quarterly basis. The locations monitored and the monitoring results should be recorded. Survey and monitoring data should be audited periodically by management a t a frequency commensurate with the scope of the program (eg., for uranium mills, a t least annually). As noted in Section 2, data review by outside observers may also be appropriate a t a frequency dependent on the complexity of the radiation safety program.

5.3.2 Personal Monitoring-External Evaluation of radiation doses to individuals is an essential component of the radiation monitoring program. It is important to recognize that exposures can be of two types, external and internal, and that it is the sum of these two types that must be assessed against the dose limits. Individual exposures to external penetrating radiation can be estimated by: (1) assigning personal monitoring devices to individuals,

5.3 EXTERNALRADIATION

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(2) using passive integrating devices as area monitors and apply-

ing suitable occupancy data to estimate cumulative doses for individuals, or (3) calculating individual doses from external radiation survey results and a record of time spent by the worker in the various areas. In general, external radiation surveys should be performed and individual doses estimated to determine the need for personal dosimeters. Personal monitoring devices should be considered for individuals likely to receive doses in excess of ten percent of the occupational limit. As a simplified rule, personal monitoring should be provided when gamma radiation levels exceed 5 p,Gy per h. Where occupancy is very low, the threshold for providing personal monitoring can be proportionally higher. In addition to providing documentationof the individual's cumulative dose, personal monitoring results may reveal developing trends or unusual practices and may be used to verify that exposures are consistent with the facility's policies and the individual's work assignment. Personal monitoring devices include TLD badges and film badges. TLD badges are currently the preferred method of personal dosimetry in mines and mills since film badges are more sensitive to harsh environmental conditions. Difficulties which may be encountered with personal monitoring devices include loss of the device; inadequate protection from water, mud and chemicals, and contamination by radon progeny. Personal monitoring devices are vulnerable to high humidity and dust and should be protected. This is especially true for films, which, if used, should be sealed in plastic envelopes. The supplier of the personal monitoring service should, and may be required by regulation to successfully, participate in a personal monitoring accreditation program such as the National Voluntary Laboratory Accreditation Program for personnel dosimetry (Berger, 1985).Accreditationwill verify the validity and enhance the credibility of the results. Additional assurance of validity is available if the supplier successfully uses other quality assurance techniques (e.g., routine analysis of dosimeters exposed to known radiation levels). The length of time the monitoring device is w o n before evaluation should be determined on the basis of the magnitude of the possible exposure, the need for information to control future exposures, the capability of the device to integrate exposure over long periods of time without loss of information, and the need to ensure that employees still have and use their monitoring devices. A monthly exchange schedule is recommended initially until actual exposure leveIs are

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verified. Schiager and Johnson (1981) suggest for uranium miners that the exchange schedule may be changed to quarterly if the monitoring results meet the following criteria: (1) less than 10 percent of the badges are lost, (2) less than 50 percent of the monthly doses exceed 1mSv, and (3) less than 10 percent of the monthly doses exceed 2 mSv. The dose as recorded by the monitoring device is subject to many sources of error, particularly with regard to a person's orientation to the radiation field, and errors up to a factor of two may occur. However, the dose determined by the primary monitoring device should be considered valid for record purposes unless one of the following has occurred: (1) the monitoring device has failed, (2) the device was exposed when not worn by the person, or (3) investigation shows that the dose was different from that recorded by the device. NCRP Report No. 101 on occupational exposures (NCRP, 198913) addresses the determination of reliable values of dose to individuals. The report includes suggestions for improving the quality of occupational data.

5.4 Long-Lived Airborne Radionuclides Two general compositions of airborne dusts containing long-lived radionuclides are of concern in the mineral extraction industries: (1) dust from the mineral prior to chemical separation and (2) dusts from operations following chemical separation. In the dusts from operations prior to chemical separations, all of the radionuclides of the uranium andlor thorium decay series may be present. Exposures to this dust may occur in the excavation process; in ore storage, crushing, grinding and drying; in other dry material handling operations in beneficiation plants; and at the input of chemical operations at mills and chemical plants. Dusts from products, by-products or wastes following chemical separations will contain compounds of uranium (such as "yellowcake") or thorium or some nonequilibrium combination of members of the uranium or thorium series. Workers may be exposed to these dusts during operations such as product drying, packaging and handling, or in the course of equipment cleaning and maintenance. Several approaches are available for analysis of air samples. Specificradionuclideanalyses provide the most dosimetrically-meaningful information. However, gross-alpha analyses provide more rapid

5.4 LONG-LIVED AIRBORNE RADIONUCLIDES

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information and are less costly. Gross-alpha analyses may prove to be the most cost-effective method for routine and early warning monitoring, provided there is direct or indirect information about the likely radionuclide composition of the airborne material. Radiometric analysis techniques are described in documents such as NCRP Report No. 58 (NCRP, 1985a). As a n alternative to radiometric analysis, samples may be analyzed chemically for uranium (eg., by fluorimetry) in the case of ores and products in which the radioactivity is primarily due to uranium or uranium with its decay series. Still another alternative is to monitor for airborne dust concentration on a routine basis after having established the typical radionuclide content of the material being handled. For many grades and types of ore, the exposure limit for dust based on the silica content is more restrictive than the exposure limit for dust based on uranium or radionuclide content. 5.4.1 Characterization of the Workplace

Area monitoring for airborne dust serves to identify areas requiring special control procedures and posting, to determine the adequacy of the ventilation and dust control program, to identify locations for which individual exposure evaluations are indicated, and to demonstrate compliance with exposure limits. Area sampling is performed with fixed or portable sampling equipment located so as to represent the atmosphere in the general area. Air sampling should be performed initially at sufficient locations and frequencies to estimate potential exposures to individual workers. In areas where frequent or continuous air contamination is likely to be a t levels more than ten percent of derived occupational exposure limits, air should be sampled continuously during periods of occupancy. Examples involving a high potential for airborne radionuclides include product drying and packaging areas. Where results indicate that no individual is likely to be exposed a t levels exceeding limits applicable to the general public, further sampling is indicated only when changed conditions are expected and when confirmatory sampling is desirable. Other areas having the potential for airborne radionuclides should be sampled periodically, consistent with the guidance of Table 5.1. In addition, regulatory requirements or license conditions may specify sampling requirements for specific facilities. 5.4.2 Personnel Exposure Assessment-Internal

For individuals potentially exposed to more than ten percent of the airborne radionuclide limit, individual exposures should be

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evaluated and recorded. Several different methods are available; use of personal samplers, calculation from breathing zone sampling or interpretation from area sampling. The most direct method of documenting individual exposures to airborne radionuclides is through the use of personal samplers. Personal samplers consist of small portable pumps worn on the person and small sampling heads worn as close as convenient to the individual's mouth and nose. These sampling heads are often worn on the lapel; hence, the systems are often referred to as lapel samplers. Another method of estimating individual exposures is through breathing zone sampling. In this type of sampling, sampling stations or portable samplers are fixed or held in a position closely representing the breathing zone of an individual at a defined work station. The duration of exposure to each of the various monitored conditions is recorded for the individual being monitored. Average concentrations or cumulative exposures are then calculated from the observed breathing zone concentrations and the respective exposure times. The least direct method of estimating individual exposures is by calculating from general area monitoring results and occupancy times. While dust concentrations in mines can generally be kept low enough to preclude the need for individual monitoring, the need for such monitoring should be examined more closely in dusty uranium and thorium product handling areas. Examples of such areas include yellowcake drying and packaging areas in uranium production. Individual monitoring should be provided during special high exposure operations such as maintenance of yellowcake drying and packaging equipment and in areas of potential release not monitored by fixedlocation samplers. Samples for individual exposure assessment, whether collected with personal samplers, breathing zone area samplers or general area samplers, should be representative of the air breathed by the worker. Sampling should be of a sufficiently long duration to represent the operation being conducted and the quantity of air sampled should be large enough to provide the required sensitivity of measurement. Care should be taken to avoid extraneous contamination of samples. 5.5 Airborne Radon and Progeny

For the purpose of exposure assessment, the concentration of airborne radon progeny is either measured directly or inferred from radon measurements and knowledge about equilibrium ratios.

5.5 AIRBORNE RADON AND PROGENY

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5.5.1 Characterization of the ~ o r k ~ l a c e - ~and ~~R Progeny n Initially, air sampling should be performed a t sufficient locations and frequencies to determine potential exposures to individual workers. Subsequently, routine surveys should be continued in accordance with the criteria in Table 5.1 in all areas having the potential for elevated airborne concentrations of radon progeny. In general, concentrations in uranium mines will dictate more frequent sampling intervals, while concentrations in other mines will usually be lower and routine sampling, if required, can typically be a t less frequent intervals. In mills, radon and radon progeny constitute a secondary source of exposure confined mainly to ore storage, processing and handling areas. General air samples should be collected in the vicinity of these operations a t a frequency depending upon the magnitude of the observed airborne concentrations. A variety of methods has been developed for radon and radon progeny sampling. Most are grab sampling (short-term) methods, although continuous monitoring methods are under development. Monitoring methods have been described by the ANSI (1973)and the ICRP (1977).Measurement techniques and applications to particular situations are addressed in some detail in NCRP Report No. 97 The method used should have a sensitivity sufficient (NCRP, 1988~). to make the decisions prompted by Table 5.1. Since radon concentrations may vary by a n order of magnitude over short periods of time, single measurements should be interpreted with caution. A series of grab samples or continuous radon monitoring is usually required to establish average annual levels. Grab sampling can be greatly facilitated by the use of prompt measurement methods (i.e., "instant working-level meters") (Schiager et al., 1981). If routine monitoring indicates unusually high concentrations for a particular area, the measurement should be repeated, the source of the increase identified, corrective action taken and additional monitoring performed to verify the effectiveness of the corrective measures.

5.5.2 Personnel Exposure Assessment--222Rn and Progeny Individual exposures should be evaluated and recorded for personnel likely to be exposed to more than ten percent of the occupational limit and for persons occupying areas in which the concentration exceeds the derived air concentration. [Monitoring is required by

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regulation in facilities licensed by the Nuclear Regulatory Commission (NRC) and some states when exposures are likely to exceed a specified fraction, for example, 10 or 25 percent of the limit (see Section 71.1 This can be done by computing time-weighted exposures on the basis of average measured concentrations from area sampling and occupancy times. To assess personnel exposures properly, air samples should be representative of worker exposure and be taken near locations where workers are most often present. At the present time there are no personal radon or radon progeny monitoring systems in use in the United States. However, a number of passive and active systems are in the conceptual stage, under development or being tested. Such systems have been adopted in other parts of the world (Schiager et al., 1981; NCRP, 1988c; Brown, 1992) and show promise for the future.

5.5.3 Radon-220 and Progeny In general, the measurement concepts presented for "'Rn and progeny apply to "ORn and progeny, although there is less practical experience upon which to base specific procedures. Since there are significant differences in the half-lives (see Glossary) between the two radon isotopes and between the controlling members of the two progeny series, mixtures of "ORn and 222Rnor of their respective progeny can be resolved in grab samples on the basis of a series of gross radioactivity counts a t appropriately selected time intervals after sample collection. Methods for monitoring for "ORn and progeny products and 222Rn/220Rn progeny mixtures have been published by Rock (1975), the IAEA (1976a), Khan and Phillips (1986) and the NCRP (19884.

5.5.4 Monitoring for Control Purposes In addition to monitoring for assessment of exposure to personnel, measurement of airborne radon and progeny is conducted in underground mines for control purposes, i.e., for evaluation ofthe effectiveness of ventilation and other mechanical radon-control measures. Radon measurements indicate the degree of control over the gaseous source and the potential for airborne progeny. The ratio, radon progeny concentration to radon concentration, is useful in identifying the location of the major sources of contamination relative to the make-up air and the point of observation.

5.6 SURFACE CONTAMlNATlON

47

5.6 Surface Contamination The need for, and the frequency of, contamination monitoring should be governed by the assay of the material being handled, the potential of the operation for producing contamination, the use of the area involved and the results of actual monitoring. A surface contamination monitoring program should be instituted for areas where the material being handled contains significant concentrations of uranium, radium or thorium and the operation being performed is such that surface contamination could result (potential contamination areas). Total surface contamination can be adequately monitored with the typical alpha-specific survey probe. Surveys for removable contamination should be performed by "smear" or "wipe" tests with laboratory counting of the smear or wipe. Some forms of uranium product are readily visible because of a bright yellow color. In areas where this clearly identifiable form is known to be the only radioactive material, qualitative contamination surveys can be performed by visual inspection. Beta-gamma survey probes may be appropriate for those cases where nuclides decaying by beta decay (e.g., 'lOPb; see Appendix A) are believed to be predominant among the radionuclides on the contaminated surface. In mines and mills, ore dust settles on surfaces in the normal course of operations. In dry locations, uranium or thorium ore dust may be of concern as a possible source of air contamination by means of resuspension. However, this is not usually a problem because of the very low specific activity of the ore, and monitoring of surface contamination is not required in such areas. Surface contamination may be a problem in mills where uranium or thorium concentrates are handled, such as in the precipitation, filtration, weighing and packing areas. A surface contamination monitoring program for such potential contamination areas should include monitoring of areas, persons and objects such as tools, equipment and packages for shipment. 5.6.1 Area Monitoring Area surveys for surface contamination should be performed on a regular basis in potential contamination areas. Surveys for removable surface contamination should be performed in nearby rooms where work with radioactive materials is not performed but where surface contamination should be kept at low levels. These include areas such as lunch rooms, change rooms, control rooms, maintenance shops, analytical laboratories and offices.

48

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5. MONITORING OF OCCUPATIONAL EXPOSURE

5.6.2 Monitoring of Personnel Personnel should be monitored with a survey probe for skin and clothing contamination as they leave an area designated as a potential contamination area. Surveys for alpha and for beta-gamma detection may be appropriate for the personal monitoring, depending on the work being performed and the material being handled. 5.6.3 Monitoring Other Items

Respirator facepieces and hoods should be surveyed for alpha contamination before being reused. Items such as tools, equipment and vehicles being removed from a potential contamination area should be monitored to preclude inadvertent radiation exposure in uncontrolled areas. Potentially contaminated components that have been removed h m the mineral processing equipment should be monitored before being released for maintenance or for use outside the contamination area. Packages of radioactive product should be surveyed for contamination before shipment. Packages having contamination at levels that exceed facility administrative limits or the relevant transportation standards (Section 7) should be cleaned and resurveyed prior to shipment. Tools, equipment, vehicles and packages of product should be cleaned to contamination levels which are ALARA above background before they are released for use outside the area controlled for radiation protection purposes. Further, facility administrative limits should be established to identify surface contamination levels above which the facility operator will maintain contamination controls and ensure the materials remain within the controlled area. These limits are established considering potential equivalent doses to workers or members of the public resultant from contact with the contamination and the practicability of detection of the contamination. NCRP Report No. 116 (NCRP, 1993)presents recommendations on limits for exposure to ionizing radiation. That report also describes a "negligible individual dose" that is an annual effective dose of 0.01 mSv per y per source or practice. A dose a t this level can be dismissed. Methods such a s those described by Dixon (1984) may be used to derive contamination limits within the dose limiting recommendations described by NCRP. Such limits generally take the form of exposure-rate limits on radiation fields around the materials and limits on contamination removable from the materials. Regulatory agencies have also established guidance and limits for release of materials for unrestricted use, for example, in establishing facility decommissioning criteria (see Section 7).

5.7

5.7

BIOASSAY

/

49

Bioassay

Bioassay is the determination of the kind, quantity, location and1 or retention of radionuclides in the body. A comprehensive discussion of bioassay is presented in NCRP Report No. 87 (NCRP, 1987b). Bioassay programs assess the adequacy of control over intake of radionuclides, detect any accumulation of radionuclides in the body, and provide the basis for calculating radiation dose from radionuclides in the body. Bioassay techniques also confirm the effectiveness of the airborne measurement program. Some specific objectives of bioassay include: (1) monitoring any radionuclide accumulation that occurs as the result of continued low-level exposure, (2) detecting any significant deposition events, (3) verifying the effectiveness of respiratory equipment when personnel are exposed to high concentrations of airborne radionuclides, and (4) assessing the consequences of known or suspected internal exposure events such as accidents or significant transients in airborne radionuclide concentrations. Bioassay programs should be considered for workers in areas where unsealed radioactive materials are handled and there is the possibility for inhalation of airborne radionuclides or for ingestion of loose radioactive contamination. Uranium bioassay programs should be initiated for workers in uranium product areas (i-e.,yellowcake workers) and for any other workers with comparable potential for radionuclide intake. 5.7.1 Bioassay Methods

Bioassay methods fall into two categories, direct and indirect. In the direct or in vivo method, a whole- or partial-body counter is used to estimate directly the body burden or organ burden of radionuclides. In this method, the emission of photons from internally deposited radionuclides is measured by one or more radiation detectors located external to the body. Indirect bioassay consists of the in vitro measurement of radioactivity in material excreted or removed from the body. In this case, body burdens are estimated from the in vitro results by calculations using accepted biological excretion models. Urinalysis is the primary bioassay procedure for determining whether intake of uranium has occurred. Uranium in the urine represents excretion of material that has been absorbed into the body. In addition, fecal analysis may be used on occasion to detect

50

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5. MONITORING OF OCCUPATIONAL EXPOSURE

the passage through the gastrointestinal tract of unabsorbed material from recent ingestion or from swallowing following inhalation. For many exposure circumstances, in vivo measurement is useful and is often the bioassay method of choice for assessing internal deposition of radionuclides which emit penetrating gamma-rays with a reasonably high abundance. Measurement of gamma-rays emitted from the chest region (lung counting) has been employed to detect the possible inhalation and retention in the lung of uranium in chemical forms with low solubility (long clearance time). However, because of the low sensitivity for uranium and the attendant requirements for low background radiation levels and trained personnel, this method is not practical for routine programs in mines and mills but may have applicability for special investigations. Fisher and Stoetzel (1983) recommend that chest counting for uranium be performed only a t institutions with fixed equipment, graded shielding and highly qualified personnel capable of measuring and interpreting deposition of uranium in the lungs. For special investigations, other radionuclides, such as 210Pb(Laurer et al., 1993) are identified using in vivo bioassay.

5.7.2 Bioassay Program Content Bioassay programs should be consistent with the properties of the radioactive material of concern and the nature of the operations in the facility. Procedures should have a sufficient sensitivity for the decisions required. The program should include an adequate quality assurance/quality control component as indicated in Section 2.6. In general, the bioassay program should provide for: (1) routine, periodic measurements, (2) specially scheduled, post-exposure measurements and (3) follow-up studies. The program should also include initial and close-out measurements. An initial or baseline measurement should be performed for all personnel who will be working in areas requiring a routine bioassay program and for persons in those occupations andlor areas in which there is the potential for requiring special studies. This identifies pre-existing body burdens that may have been received prior to employment in the facility and provides a baseline value against which to compare future results. Similarly, when a worker will no longer work in areas where bioassay is required, close-out measurements should be made to determine and document any accumulation of radioactive material in the body.

5.7 BIOASSAY

5.7.3

1

51

Routine Bioassay

Routine bioassay programs involving periodic measurements are performed to monitor radionuclide accumulation from chronic, lowlevel exposure and to detect significant deposition events from singleexposure episodes. A routine program involving periodic measurements should be instituted for individuals who work in areas in which ambient concentrations of airborne radionuclides exceed ten percent of the limit. NCRP Report No. 87 (NCRP, 1987b) provides additional information for defining participation criteria for the bioassay program. In order to detect internal deposition, measurement should be a t a frequency such that depositions of concern are successfully detected with the selected measurement technique before becoming undetectable due to elimination of the radionuclide from the body. In NCRP Report No. 87 (NCRP, 1987b)procedures are illustrated for establishing the maximum interval between specimen collections for detection of a single intake that will produce a transient deposition a t a stated level of interest. In general, more frequent sampling should be employed for personnel working in those areas with the highest potential for radionuclide intake, for example, semi-monthly to monthly bioassays should be considered for workers in yellowcake precipitation and packaging. 5.7.4

Post-Exposure and Follow-Up Measurements

Post-exposure measurements should be performed followingknown or suspected intakes of radionuclides. For example, a bioassay should be performed for any individual who: (1) shows evidence of significant skin contamination, especially around or in the nose and mouth; (2) has been exposed in excess of the derived weekly intake limit [i.e., >I150 of t h e a n n u a l reference level of intake (see Glossary)], with or without respiratory protection; or (3) has been involved in a n incident that caused a possible accidental intake. The time of sample collection is a n important variable in excretion analysis following exposure events. For example, for the greatest sensitivity in detecting recent intakes of uranium in the oxide form, urine samples should be collected during the time interval from 48 to 96 hours after the potential exposure. For uranium in more soluble forms, collection should begin sooner. Follow-up studies should be performed if the results of routine or specially scheduled bioassays indicate the presence of an organ or

52

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5. MONITORING OF OCCUPATIONAL EXPOSURE

body burden corresponding to greater than ten percent of an annual reference level of intake. These follow-up measurements serve to verify the initial analysis and to determine the actual turnover function for a more refined dosimetric assessment of the intake. In the event that restrictions were placed on a worker because of bioassay results, the follow-up measurements also indicate when the restrictionscan be removed. Follow-up measurements should be performed with sufficient frequency to define the turnover function before elimination of the material from the body renders the levels undetectable.

6. Effluent Monitoring and Environmental Surveillance Operators in the mineral extraction industries should monitor releases of radioactive materials to the environment to determine radiation exposures, define the extent of contamination, determine compliance with government regulations and corporate guidelines, and provide a basis for assessing the efficiency of effluent control systems. In this Section, environmental and effluent monitoring programs are presented, based on radiation protection principles. Additional monitoring may be required or prudent a t specific facilities based on site-specific conditions. Criteria are presented in this Section to determine the need for limited or routine environmental monitoring. However, before a mineral extraction facility is built, baseline environmental samples should be collected.Those samples are used to document the radionuclide concentrations present prior to the operation of the facility and to segregate facility-inducedconditions from pre-existing conditions, including those from other facilities. The analysis of air, water, soil and vegetation samples is recommended for radionuclides and for chemical elements that have a known or potential influence on the bioaccumulation of radionuclides in plants or animal^.^ This is important for adequate evaluation of potential impacts since the uptake of nonessential radionuclides may be strongly influenced by the abundance or scarcity of an essential chemical element analog (Vanderploeg et al., 1975). For example, if radium in water is an effluent from the facility, the analysis of calcium in the receiving waters or soil is advisable because the uptake of radium by plants or animals has been observed to be influenced by the availability of calcium (Hansen et al., 1960;Lindeken and Coles, 1978). The program for the collection of baseline data typically requires compilation of up to one year's data for comparison with the data collected during operations. If seasonal trends are observed in the 4SeeSection 6.3.2for additional information on baseline program design and compatibility with operational monitoring.

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6. MONITORING AND ENVIRONMENTAL SURVEILLANCE

baseline concentrations, those concentrations can be used to adjust the data collected during operations to obtain a measure of facility effluents present in the environment.

6.1 Environmental Pathways

Figure 6.1 presents the principal potential radiological exposure pathways to humans from mineral extraction operations. (Other pathways may exist at some facilities, for example, release from tailings impoundments to surface water.) The eflluents from the sources on the left side of the figure may enter the atmosphere, surface water or groundwater where dilution, dispersion, concentration, reconcentration, deposition or resuspensionmay occur; radioactive ingrowth and decay may occur as well. Ultimately, a portion of the radioactive materials may be transferred to humans from the environment. Once the materials are inside the human body, radiological exposure to the whole body and to individual organs occurs. The degree of exposure is dependent on the individual radionuclide,

DISPERSION MEDIUM

SOURCE

-

b

TAILINGS

$ EXTRACTION - FAClLrrY (MILL) a

*

PRODUCT

t ORE TRANSPORT

*

I

-

--+

PLANT

AIR

INHALATION EXrERNAL RADlATlON SOILS

UPTAKE

CROPS LIVESTOCK MEAT MILK UPTAKE

v

t

SURFACE WATER

MINE

RUNOFF + COLLECTlON PONDS

IVLMJSPHERE

-

PATHWAY

WATER FISH

It GROUND WATER

t

-f

-

,N

SEDIMENTS EXTERNAL RADIATION WATER

+

s

Fig. 6.1. Principal radiological exposure pathways to humans from mineral extraction industries.

6.2 EFFLUENT MONITORING

1

55

its chemical and physical form, its concentration and how it is metabolized. Further details can be found in NCRP Report No. 76 (NCRP, 1984b) and IAEA Safety Series Nos. 77 and 90 ( M A , 1986; 1989).

6.2 Effluent Monitoring 6.2.1 Effluent Monitoring Objectives

Specific objectives to be met by the design of each monitoring and surveillance program include: (1) collection of representative samples of effluents a t a frequency that will allow corrective actions prior to any significant environmental contamination or radiological exposures. (2) collection of representative samples a t locations that allow assessment of environmental impacts. Usually such locations are a t or near the point of release to the environment, for example, in a mill scrubber stack or a t the discharge pipe from a mine water treatment plant. Detection of radionuclides is easier a t such locations because dilution and dispersion in the environment have not yet occurred. (3) analysis of t h e collected data for trends of increasing or decreasing concentrations as a function of the time of collection to determine whether changes i n monitoring programs or effluent controls are warranted. Trends should be reviewed annually or a t a frequency which correlates with major repairs or changes i n equipment t h a t could impact effluent concentrations. (4) assurance to the public that all significant impacts resulting from the operation of the facility are being monitored and analyzed. (5) validation of the measured values by documentation of the sampling and analytical procedures and quality assurance measures. 6.2.2 Program Design

The necessity for effluent monitoring should be assessed initially and a t least when there are changes in equipment that could affect effluent concentrations. Facility operators need to know or determine the radionuclide content and concentration in each process circuit, points within the circuits where radionuclides can concentrate,

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6. MONITORING AND ENVIRONMENTAL SURVEILLANCE

points of potential releases to the environment and the radionuclide concentrations that could be or are present in effluent streams. Consideration should be given to the potential for release of radionuclides to the environment via pathways such as sanitary or storm sewers in addition to the direct process-circuit pathways. Those data are used to provide an estimate of effluent concentrations in the environment5,an estimate of the effective dose (radiation dose from effluents released to the environment), and a list of radioactive materials to be measured in the effluent and environmental monitoring programs (see Section 6.3.2). The NRC has reported on effluents from uranium mines and mills and has estimated doses to residents near those facilities from those effluents (NRC, 1980). Such information may supplement site-specific information in defining effluent program design. Also, guidance is emerging on approaches to demonstrate compliance with EPA emissions standards (Rhodes, 1992). Effluent monitoring should continue in effect when potential doses are greater than ten percent of the nonoccupational dose limit^.^ When ongoing effluent monitoring is warranted, the interpretation of data must be considered. Large area sources, multiple release points and short sampling times relative to long effluentdischarge time periods each introduce complexities in interpretation of environmental data. The monitoring program design should minimize uncertainties of data interpretation. Site-specific validation of environmental assessment models is recommended as a complement to the effluent monitoring program design. To ensure meaningful evaluation of the directional transport and dispersion portions of the model, assistance from qualified individuals may be required in designing an effective validation test. 6.2.3 Air Monitoring

Airborne emissions from underground mines occur at the mine ventilation shafts. Usually ventilation air is brought into the mines through the equipment-ore haulage shaft, routed to the working areas of the mine and exhausted through several ventilation shafts 6Calculational models are often used to estimate environmental concentrations resultingfrom concentrationsin effluent. Use of such models typically requires knowledge of site-area meteorological data (eg., wind speed, wind directions and temperatures). Such data may be obtained from meteorological towers constructed on-site for that purpose or from other information sources believed to be representative of sitearea conditions. 6Nonoccupationaldose limits are presented in NCRP Report No. 116 (NCRP, 1993) or are established by regulation (see Section 7.)

6.2 EFFLUENT MONITORING

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located away from the main haulage shaft. The extensive atmospheric dispersion induced by the high velocity discharge from those ventilation shafts normally results in site-boundary concentrations which are indistinguishable from background levels. Thus, a monitoring program for the exhaust air would be appropriate for underground mining operations when members of the general public live nearby. For operations remote from the general public, a monitoring program for exhaust air is not recommended. Monitoring of airborne effluents from open pit mines is not usually warranted because of the large surface area available for release of emuents and the large atmospheric dispersion associated with open pit mines. Again, monitoring would usually result only in background-level measurements and confirm the limited potential for exposure of members of the public. The highest concentrations of airborne effluents from mills are usually from the mill exhaust stacks. The stacks most likely requiring monitoring are those that release effluents which could result in significant exposures to humans via direct inhalation [NCRP (1993) presents recommendations on limits for exposure to ionizing radiation]. Such exhaust stacks include those from the yellowcake dust collector a t uranium mills. Isokinetic sampling through monitoring ports installed in the side of such stacks is recommended (EPA, 1985). For stack effluents that produce lower potential exposures, less precise and simpler measurements, such as grab sampling using portable equipment, are useful in determining whether the stacks contribute significantly to the airborne radionuclide emissions from the facility. The screening models discussed in the next section will help determine the significance of the stack emissions. The frequency of monitoring is a function of the potential for a release from the facility, the potential impact of the effluent on people and the environment and the variability in release rate over time (including the reliability of the effluent control system). When exposure or concentration limits are being approached or a significant potential for a large release exists, the sampling should be done frequently, possibly continuously. As a n alternative to a statistical evaluation of the sampling frequency and total number of samples necessary for specific circumstances, Table 6.1 is presented as a guide for a stack sampling program a t a uranium mill, which may represent a worst-case situation for the minerals extraction industry. Other facilities, which may have smaller quantities of radionuclides or have a lower potential for a large release, may require less frequent sampling. Sufficient sample volumes need to be collected to meet the lower limits of detection (LLD) of the analytical laboratory and as specified in the applicable regulations. Where practica-

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TABLE6.1-Stack

monitoring frequencies and sample types for a typical uranium mill.

Effluent

Materials

Potential

Operational

Stack

Special

control

controlled

exposure

check

monitoring

monitoring

frequencf

fkquency

Wet scrubber

equipment

yellowcake

high

hourly or continuous

quarterly

isokinetic

requirements

Bag dust collector

U ore dust

moderate

monthly

semiannually

-

-

Mist eliminator

acid mists low annually one time from U ore only leach tanks " An operational check is a test of the effluent control equipment to ensure that it is functioning as designed. Selection of the type of check depends on the design and reliability of the equipment and the need to ensure that effluent controls are at least as effective as assumed in environmental asseswnent modeling.

ble, multiple samples should be collected and analyzed, to provide a higher level of confidence in the results [see NRC Regulatory Guide 8.30 (NRC, 1983a) and NCRP Report No. 58 (NCRP, 1985a)l. Water Monitoring Water containing radionuclides may potentially be released to constructed ponds, surface waters or waterways. The subsequent fate of the radionuclides is indicated by the exposure pathways presented in Figure 6.1. Representative grab samples of water emuent should be obtained a t the point of release from mineral extraction facilities. That point is often the pipe releasing water from the water settling ponds or from the water treatment plant. The analyses of both suspended and dissolved constituents are recommended where both water fractions can transport radionuclides to people or the environment. The frequency of sampling may be determined as described in Section 6.2.3. For facilities with reasonably stable release rates, monthly composite samples consisting of a t least three samples may be sufficient. Automatic, continuous, water samplers are available and can facilitate the sampling tasks [see NRC (1983b)for suggested LLD values].

6.2.4

6.3

Environmental Surveillance

6.3.1 Environmental Monitoring Objectives

The objectives of an environmental monitoring program for radiation protection are: (1) to assess the radiation and radionuclide levels in the preoperational environment,

6.3 ENVIRONMENTAL SURVEILLANCE

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59

(2) to detect changes in levels to verify the adequacy of effluent assessment and control programs and to evaluate loss of control, (3) to assess actual or potential exposure of members of the public from facility effluents and from direct (y-ray) radiation from the facility, (4) to determine the fate of contaminants released to the environment, and (5) to demonstrate compliance with applicable regulations or other legal requirements. In achieving these objectives, the facility operator is able to provide assurance to the public that potentially significant impacts are being monitored and analyzed. 6.3.2 Program Design

In 1984, the NCRP published a statement concerning the proposals made by EPA under 40 CFR 61 on the control of air emissions of radionuclides (NCRP, 1984d). This statement reviewed existing limits in published NCRP reports and works in progress and emphasized that for continuous exposure of an individual in the population a n annual effective dose of 1mSv should be limiting. The statement went on to say that whenever the potential exists for a n individual to exceed 25 percent of the limit for whole body exposure from any single site, the site operator should be required to assure that the annual whole-body effective dose of the maximally exposed individual from all sources would not exceed 1mSv on a continuous basis. The operator of a mineral extraction facility needs to determine appropriate means to assess exposure of individuals and more specifically the need to conduct environmental radiation monitoring prior to and/or during facility operations as a means of exposure assessment. Screening models may be used to help operators determine if a n environmental monitoring program should be conducted during facility operations. Note that exceeding a screening level does not imply any noncompliance with regulations or limits. The screening model set out in NCRP Commentary No. 3 (NCRP, 1989b) is limited to routine operational releases of specific radionuclides to the atmosphere over a period of one year. If the calculated radiation doses to individuals as determined using the screening model do not exceed the nonoccupational dose limits7a t any succes7Nonoccupationaldose limits are presented in NCRP Report No. 116 (NCRP, 1993) or are established by regulation (see Section 7).

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sive level in the model, environmental monitoring during operations is not considered necessary. Conversely, if the calculated screening value exceeds the applicable nonoccupational dose limit at the final level of screening, then an environmental monitoring program is recommended. A similar methodology is described by the EPA (1989a) to demonstrate compliance with the rules of 40 CFR 61 (EPA, 1989b). When airborne radioactive materials other than those presented in the screening model (NCRP, 1989b)are released from the facility or when radioactive materials are released into a water pathway, the following screening mechanisms may be used to determine the necessity of environmental monitoring during operations. (1) If the concentrations of radionuclides in water or air as measured at the discharge points from the facility are calculated to result in a committed effective dose of less than 1 mSv in a year, environmental monitoring is not necessary since distance between the discharge points and the location a t which an intake would usually occur dilutes the effuents to concentrations below which environmental monitoring is recommended. (2) In the unusual case where the public has access to air or water directly at the discharge point, initiation of environmental monitoring would be appropriate because the dilution factor mentioned above would not be applicable. (3) If the concentrations measured at the discharge points are calculated to result in an annual dose higher than the 1mSv committed effective dose limit, environmental monitoring is recommended. This method of determining the necessity of environmental monitoring is a t least four times more restrictive than the first level of screening for atmospheric releases set out in NCRP Commentary No. 3 (NCRP, 198913).The NCRP plans to publish a comprehensive screening model report for more than 800 radionuclides, and for both air and water pathways. For those facilities for which environmental monitoring during operations is recommended, baseline (preoperational) environmental monitoring is also recommended. Such monitoring enables appropriate consideration to be given to preexisting conditions on and near the site of the facility. If the screening results in a determination that ongoing environmental monitoring is not necessary, the facility operator should still evaluate the desimbility of conducting environmental monitoring prior to andlor during operations. Objectives of environmental monitoring programs remain valid even when the probable upper limit of exposure has been reasonably estimated and compliance with

6.3 ENVIRONMENTAL SURVEILLANCE

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regulations is reasonably assured. Corporate philosophy regarding scientific investigation and provision of additional levels of assurance of facility safety may lead to the decision to perform monitoring. For example, monitoring may be determined to be prudent to verify the applicability of the screening model to site-specific conditions. Similarly, performance of monitoring may be determined to be prudent to ensure preexisting conditions are properly considered in facility operations. If the decision has been made to conduct environmental monitoring, samples representative of environmental concentrations need to be collected. Sampling usually extends over the duration of at least one year to demonstrate possible seasonal variations. If the potential radiation doses to individuals attributable to facility effluents and calculated from the results of environmental sampling are ten percent or less of the nonoccupational dose limits, continued environmental monitoring may not be needed. Such a decision should be reevaluated if changes in the extraction facility are made which would increase the actual or potential effluents or if more than one source exists in the area. Then, environmental monitoring should continue consistent with the criteria described above and with corporate policy. The environmental monitoring should be for those radioactive materials in each major pathway that would be expected to contribute substantially to the calculated radiation dose. Effluent monitoring, such as stack or water discharge monitoring, continues when potential doses are greater than ten percent of the nonoccupational dose limits. Then, environmental monitoring should also continue, to provide assurance that corrective actions could be implemented before the nonoccupational dose limits are exceeded. In the design of a n environmental monitoring program, the baseline environmental monitoring program and the operational program should be made as compatible with one another as possible. For example, using as many of the same sampling locations as possible allows direct comparison of the data collected to assess conditions before and after facility operations commence. Sampling of the following types of media, representative of major pathways in environmental monitoring programs, is recommended in the baseline and the operational monitoring program: (1) airborne particulate material, gases and gamma radiation near the site boundaries of the property, as determined by fencing and posted signs, and in the directions that have the highest predicted concentrations of airborne particulates; (2) airborne particulate material, gases and gamma radiation near the closest residence or structure that is occupied for a

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portion of each year and near the residence that has the highest anticipated concentrations of airborne radionuclides, eg., the nearest residence and the nearest downwind residence; (3) airborne particulate material, gases and gamma radiation at the nearest population center that could be affected by the facility; (4) water samples from the uppermost aquifer that is, or could potentially be used as, a water source and is hydrologically down gradient from major potential sources of seepage, such as tailings ponds; ( 5 ) water samples from at least one downstream surface water sampling point or from potentially affected surface water ponds; (6) human food crops, fish (if applicable) and livestock feed samples from locations where the highest airborne radionuclide concentrations are predicted; (7) soil samples from areas of the highest predicted airborne radionuclide concentrations; (8) all sample types at control locations upwind, upstream or up groundwater-flow gradient from the extraction facility and sufficiently distant so as not to be affected by facility operations, yet in the same airflow patterns, stream or aquifer as is sampled above. Analyses requested on samples from the selected locations should be based on the radioactive materials to be released from the facility and their expected contribution to radiation dose, as well as the chemicals and metals associated with the ore or extraction process, and their expected contribution to environmental impact. Consideration may also be given to infrequent screening analyses for other radionuclides, chemicals and metals, to verify that they are not found in unexpected concentrations. In the environmental monitoring program, the selection of monitoring locations is often complicated by site-specific topography, demography and the proximity of one facility to other similar facilities. For example, the presence of deep canyons or valleys near the facility can cause elevated concentrations of airborne effluents as diurnal winds move up and down the valley or canyon. If residences a r e located in those canyons or valleys, higher t h a n normally expected exposures may occur a t those locations. During emergencies, such a s large spills or releases to the environment, sampling and monitoring should begin a s soon as possible. If possible, samples should be collected as a function of both time and distance and from sample locations identified in advance of an emergency. In general, more samples should be collected and preserved,

6.3 ENVIRONMENTAL SURVEILLANCE

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if applicable, than is deemed necessary a t the time of the incident. The number of samples to be analyzed can be determined a t a later time. 6.3.3

Radon

If integrated or continuous sampling techniques are not used, radon sampling can be very time-consuming due to the numerous samples needed to characterize the natural variability of radon concentrations. Radon monitors using TLD or etched-track integration techniques are recommended to determine radon concentrations. In general, monitors using the TLD system perform better in dry climates than in wet climates. The etched-track system may require visual counting of etch marks in the detector to measure radon concentrations. Human error (fatigue) associated with the visual counting is then a factor to be included in the interpretation of results. For a discussion of radon dosimetry see NCRP Report No. 77 (NCRP, 1984a). For a detailed discussion of radon measurement techniques see NCRP Report No. 97 (NCRP, 19884. Radon flux (exhalation rate per unit area) measurements may be used to define a radon source term for a n extended source such as a tailings pile, but numerous replicate samplings in time and location are necessary to establish long-term average radon emissions. Further, the techniques used for flux measurements disturb the system being measured (NCRP, 1988~).Flux characterization may be of some value in a reclamation program or in tests using specific regulatory-agency dispersion models but is not recommended as a compon e n t of a n effluent monitoring or environmental monitoring program. 6.3.4

Radon Progeny

Radon progeny concentrations are not routinely measured in the outdoor environment because of the dilution of radon progeny that occurs outdoors. (The magnitude of the exposure received outdoors from radon progeny is only a small percentage of the potential indoor exposure levels.) Radon measurement is easier for continuous or integrated sampling, and calibration and quality control are more readily achieved for radon (versus progeny) measurements (NCRP, 1988~).Accordingly, radon progeny measurements outdoors are not considered a necessary component of an environmental monitoring program a t mineral extraction facilities.

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6.3.5 Long-Lived Airborne Radionuclides Low-volume air sampling a t approximately 0.002 m3 per s is recommended for particulate air sampling because of the low maintenance requirements of the samplers as compared to that for highvolume air samplers operated at approximately 0.02 m3per s. Each unit should be equipped with an air flow regulator and an elapsed time meter which allows determination of the volume of air actually sampled even if power failures occur. If particle sizing is needed, sampling equipment appropriate for that purpose would need to be used.

6.3.6 Soil and Vegetation Terrestrial and other types of environmental sampling are addressed in NCRP Report No. 50 (NCRP, 1976). The need for soil and vegetation sampling is usually site-specific to each mining location and is based on the potential exposure pathways that are identified (Figure 6.1). For example, if cows graze near the facility, sampling grasses in the predominant wind direction may be appropriate. One factor of importance is that the sample size for both soil and vegetation sampling should be large enough to collect sufficient activity for analysis. See NRC Regulatory Guide 8.30 (NRC, 1983a) for the calculation of the LLD based on sample size.

6.3.7 Water Sampling of water both on the surface and within the ground requires the collection of nonstagnant water samples. If samples are pumped, as from a well, the electrical conductivity of the water should have stabilized prior to sample collection to acquire a representative groundwater sample. Preservation of the samples is essential to prevent plateout of radionuclides like "'Th, 226Ra,'lOPb and 'loPo on the walls of the container (Korte and Kearl, 1984). Care should be taken to ensure that the preservation technique does not result in dissolution of the suspended component of the radionuclide concentrations. Surface water should be analyzed for both suspended and dissolved radionuclides, whereas groundwater normally should be analyzed for only the dissolved fraction. That fraction acts as the transfer medium of radioactive materials to plants, animals and man, and its analysis predicts more closely the potential for exposure through ingestion pathways. (In those cases where unfiltered water is consumed, analysis of the suspended fraction may be appropriate.)

6.3 ENVIRONMENTAL SURVEILLANCE

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Chemical analyses of water samples are often included in groundwater monitoring programs to predict the movement of radionuclides through aquifers. The analysis of relatively mobile groundwater constituents, such as chloride, sulfate and total dissolved solids, can be used to determine the rate and direction of the migration of a seepage plume.

6.3.8 External Radiation When higher-than-background gamma-ray levels are present in the environment around a facility, and these levels can be attributed to the facility, gamma monitoring is recommended. TLDs are recommended for gamma detection and are usually placed in circular patterns around the radiation source at the facility. The TLDs should be exchanged every three to six months. That frequency is sufficient to allow detection of greater than the lower limit of detection for gamma radiation yet guards against the physical loss of detectors integrating exposures over long time periods.

7. Guidelines, Standards and Regulations

7.1 General

Radiation protection guidance appears in a variety of forms and originates a t a variety of levels. Such guidance includes: (1) reviews and assessments of radiation levels, exposures, doses and effects; (2) scientific recommendations for limiting radiation dose and for good practice; (3) consensus standards developed outside the governmental framework; (4) policy recommendations; and (5) rules and regulations of individual regulatory agencies. Review, guidance and standards appear at the international level; review, guidance, policy, standards, and rules and regulations are developed by appropriate entities at the national level; and states and occasionally local governments also have rules and regulations applicable to radiation. Individual facilities, in turn, should develop specific radiation protection policies, internal rules and operating procedures that are consistent with local needs and conditions, incorporate good radiation protection practices, reflect the scientific recommendations for radiation protection, and ensure compliance with regulatory requirements.' Because radiation guidance, standards and regulations applicable to the mineral extraction industries can be expected to change and evolve, those responsible for radiation protection in the mineral

'In this Section of the Report, specific numerical limits are not stated. This choice reflects the diversity of limits applicable to the various facility types and locations and ensures that this Report does not become quickly outdated due to changes in specific numerical limits.

7.2 SOURCES OF GUIDANCE AND STANDARDS

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extraction industry a r e advised to keep apprised of current developments.

7.2

Sources of Guidance and Standards

7.2.1 Scientific Recommendations

On the international level, the ICRP recommends basic radiation principles, practices and numerical limits. The IAEA recommends health and safety measures to those receiving Agency assistance and promotes the international adoption of consistent regulations in areas such as the transportation of radioactive materials. In the United States, recommendations dealing with many aspects of radiation protection are provided by the NCRP.

7.2.2 Consensus Standards Consensus standards represent the collective opinion of users, manufacturers and professionals in a particular area. Standards organizations include the International Standards Organization,the American National Standards Institute and the AmericanSociety for Testing and Materials which promote the development of consensus standards in a number of areas, including radiation protection, radiation measurement and other related topics. Consensus standards, once issued, become operational guidelines for radiation use or monitoring and may even be adopted by reference in the guides of regulatory agencies. 7.2.3 Federal Guidance and Policy

Responsibility for radiation regulation at the federal level in the United States is distributed among a number of agencies with different orientations. The President has the responsibility to establish federal radiation policy as recommended by the Administrator of the EPA. This responsibility was originally vested in the Federal Radiation Council from its formation in 1959 until 1970, when the functions were transferred to the EPA. The EPA is also responsible for generally applicable environmental standards. The National Institute for Occupational Safety and Health (NIOSH) in the Department of Health and Human Services has the

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responsibility for developing recommended occupational safety and health standards and forwarding them to the Department of Labor for consideration for rule making. NIOSH is engaged in developing a comprehensive ionizing radiation standard recommendation for underground and surface mines.

7.2.4 Rules and Regulations

The various agencies responsible for regulating radiation at the federal level in the United States derive authority under a variety of different laws. Regulations of these federal agencies may be found in the Code of Federal Regulations (CFR) under the prefix number or "Title" assigned to the respective agency. Regulatory agencies, areas of jurisdiction and regulation citations are summarized in Table 7.1. A summary document for federal standards has been published by Oak Ridge Associated Universities (Mills et al., 1989). The U.S. Nuclear Regulatory Commission (NRC) regulates the production, distribution, utilization, and disposal of source, byproduct and special nuclear material. In 1979, the definition of by-product material was expanded to include radioactivity contained in uranium and thorium processing and extraction wastes. Thus, occupational and environmental radiation exposure aspects of uranium and thorium recovery from ores, processing and utilization, and the associated tailings, are regulated by the NRC, except for those circumstances where authority has been transferred by agreement to a state. In addition to regulations, the NRC has issued a number ofRegulatory Guides and policy statements describing examples of methods acceptable to the NRC staff for implementing specific parts of the regulations. Of particular interest are Regulatory Guides in Division 3, Fuels and Materials Facilities; Division 4, Environmental and Siting; and Division 8, Occupational Health. Two agencies within the US.Department of Labor (DOL), the Occupational Safety and Health Administration (OSHA) and the Mine Safety and Health Administration (MSHA), have the authority to develop regulations relevant to occupational radiation exposure in the mineral extraction industry. The U.S. Environmental Protection Agency (EPA) has issued a variety of standards for emissions and environmental radioactivity. The U.S. Department of Transportation (DOT) prescribes standards for and regulates the packaging, labeling and transportation of radioactive materials shipped in interstate commerce by air, rail, highway and water modes of transportation.

7.2 SOURCES OF GUIDANCE AND STANDARDS

TABLE7.1-Regulatory agencies and regulations. Agency Applicability A. Occupational exposure: NRC

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Regulations

By-product, source & special nuclear materials licensees

10 CFR 20

OSHA

Workplace

29 CFR 1910.96

MSHA

Underground mines Metal and nonmetal mines

30 CFR 57 30 CFR 57 & 58, proposed

State

Radioactive materials licenses

State reeulations

B. Eflluents and the environment: NRC

Releases from licensed facilities

10 CFR 20

EPA

Uranium fuel cycle (excluding mines) Emissions of air pollutants Discharges to waters, NPDES permits"

40 CFR 190

States

40 CFR 61 40 CFR 125

Releases from licensed facilities

State regulations

NRC

Licensee general requirements Land disposal

10 CFR 20 10 CFR 61

States

Licensee general; land disposal

State reeulations

C. wastes:

D. Uranium and thorium processing sites: EPA

Standards--active facilities and inactive sites

NRC Mill tailings regulations E. Transportation:

40 CFR 192 10 CFR 40

DOT

Interstate transportation

NRC

Packaging and transportation

10 CFR 71

States

Intrastate transportation

State health and/or transportation regulations

49 CFR 170

F. Other: EPA

Drinking water

40 CFR 141

States

Drinking water; surface and ground water

State health and/or environmental regulations

'National pollutant discharge elimination system.

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States may exercise regulatory and standard-setting authority over radiation sources where this authority has not been specifically preempted by the federal government, for example, for naturally occurring radioactive materials not regulated by the federal government. In addition, the 1959 amendment to the Atomic Energy Act authorized the federal government to enter into agreements with states for the transfer to the State of regulatory authority for byproduct, source and certain quantities of special nuclear materials. As a result, about half of the states are "Agreement States." Furthermore, some states have assumed certain aspects of federal authority under the Occupational Safety and Health Act, the Safe Drinking Water Act, the Clean Water Act, the Clean Air Act, and the Resource Consewation and Recovery Act. State agencies with radiation standard-setting and regulatory authority vary, but include health, water resources, environmental, labor, nuclear energy, transportation and emergency response agencies. In establishing a radiation protection program for a particular facility, the jurisdiction in that state should be determined. Local governments may sewe as local arms of the respective state agencies or may have local ordinances and rules which are applicable to radiation in the mineral extraction industry.

7.3 Approaches to Radiation Limits Any activity which involves radiation exposure should be justified on the basis that the expected benefits excezd the predicted cost and the total radiation exposure from such justifiable activities or practices should be kept to ALARA, economic and social factors being taken into account. In addition, the equivalent dose to individuals should not exceed limits established for the particular circumstances. Basic radiation standards are expressed as limits in terms of effective dose and equivalent dose in units of sievert (Sv). Limits are specified for the whole body and for specific organs or tissues. In the case of radiation from radioactive materials taken into the body, dose is not measured directly. Consequently, dose limits are supplemented by secondary limits for individual radionuclides. Traditionally these have been presented in terms of permissible quantities of activity in the body (maximum permissible body burdens). Current recommendations are presented in terms of annual reference levels of intake (ARLIs) expressed in becquerels or in terms of committed effective dose in units of sievert.

7.4 OCCUPATIONAL EXPOSURES

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As an operational convenience for monitoring, derived limits are specified in terms of concentrations of radioactivity. Limits in air are expressed a s derived reference air concentrations in unitsgof Bq per m3. Similarly, concentration limits are also specified for liquid effluents, surface and ground water, and drinking water. In still other cases, limits which are a step further removed from effective dose and equivalent dose, such as limits for emission, surface contamination and radioactivity in soils may be established. Traditionally, numerical limits have been established for two broad categories, (1)occupational exposure of radiation workers and (2) effluents and exposure of the general public. Numerical limits for individuals of the general public usually correspond to annual doses that are one-tenth or less of the occupational limits. Limits for emissions from specific categories of sources or for exposure via specific pathways such as drinking water may correspond to even lower doses.

7.4 Occupational Exposures 7.4.1 Introduction

In certain segments of the mineral extraction industry, eg., uranium and thorium mining and milling, the principal product is radioactive, and radiation exposure is understood to be associated with the working environment. Controlled area practices such as monitoring, posting and labeling and training in radiation protection are employed, and workers are considered to be radiation workers subject to occupational exposure limits. By contrast, other mineral extraction industries present a challenging question with regard to what radiation limits should be employed for various occupational exposure situations. In those portions of these industries not devoted to uranium or thorium extraction, the radiation source is not inherently associated with the major product but is due to radioactive materials also present in the mineral deposit. Radiation exposures are incidental to the extraction of the main product and are often due to concentration in residues or by-products. Workers in these industries are not radiation workers in the traditional sense. Radiation exposures and %istorically, recommendations for concentration limits have been expressed as maximum permissible concentrations in units of pCi per em3.

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7. GUIDELINES,STANDARDSAND REGULATIONS

intakes of radionuclides should be maintained a t ALARA levels and limited to values recommended as limits for individuals of the general public unless a formal radiation control program is established and the workers are informed of relevant conditions and radiation protection practices. There may be areas or operations, such as maintenance of chemical plant equipment with radioactive residues, in which radioactivity and radiation levels are significantly higher than is typical for the balance of that industry. In such cases, it may be necessary to establish a radiation controlled area around the radiation or radioactivity source, to provide radiation protection measures, to conduct routine radiation monitoring, and to designate the involved personnel as radiation workers. Occupational exposure limits could then be applied for such personnel. 7.4.2 Recommendations Recommendations for basic radiation exposure limits have been presented in Publication 60 by the ICRP (1991) and in Report No. 116 by the NCRP (1993). Recommended limits of intake and limits of airborne concentrations of radionuclides have been presented in Publication 30 by the ICRP (1979-1988). An ARLI has been adopted by the NCRP (1993). A procedure for deriving a mixed airborne radionuclide limit for uranium ore dust is presented in ICRP Publication 24 (ICRP, 1977)and recommended limits for exposure ofworkers to airborne radon progeny are presented in ICRP Publication 32 (ICRP, 1981). NCRP Report No. 78 (NCRP, 19844 comments on standards and associated risk for exposure to radon and progeny. 7.4.3 Standards and Regulutions As summarized in Table 7.1, Section A, NRC regulations applicable to occupational exposure are contained in Title 10, Code of Federal Regulations, Part 20 (10 CFR 20), Standards for Protection Against Radiation (NRC, 1991). OSHA regulations, 29 CFR 1910.96 (OSHA, 1972) are essentially identical to the occupational exposure provisions of 10 CFR 20. MSHA health and safety standards for underground mines, 30 CFR 57, address radiation and airborne radon progeny (MSHA, 1980).MSHA has also proposed more comprehensive radiation standards for metal and nonmetal mines (MSHA, 1986). The regulations of the respective agreement states contain occupational exposure provisions that are compatible with those of NRC.

7.5 EFFLUENTS AND THE ENVIRONMENT

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Overlap between agencies is minimized by various agreements and policies. OSHA deems NRC and Agreement State licensees that are in compliance with the NRC and state regulations and license conditions to be in compliance with OSHA standards with regard to the licensed sources. OSHANSHA jurisdictions are defined by agreement, and MSHA enforces OSHA regulations in mining situations where there are no applicable MSHA regulations.

7.5 Effluents and the Environment

7.5.1 Effluents NRC radiation protection standards, Title 10 of the U.S. Code of Federal Regulations, Part 20, (10 CFR 20) (NRC, 19911, and comparable state regulations specify permissible concentrations of radionuclides in emuents released off-site from facilities licensed by that agency. Other potentially applicable NRC regulations include 10 CFR 51, "Licensing and Regulatory Policy and Procedures for Environmental Protection" (NRC, 1988). EPA standards for uranium fuel cycle activities, 40 CFR 190,limit dose to members of the public with regard to all radionuclides and exposure pathways except for radon and its progeny (EPA, 1988a). EPA standards for emissions of air pollutants, 40 CFR 61, limit dose to members of the general public from emissions from Department of Energy (DOE)facilities and from NRC and non-DOE federal facilities and limit the annual emissions of 210Pofrom elemental phosphorus plants (EPA, 1985; 1989a).These standards also specify practices for limiting 222Rnemissions from underground uranium mines (Subpart B) and from uranium mill tailings (Subpart W)(EPA, 1985; 1989a). The discharge of pollutants into United States navigable waters requires a National Pollutant Discharge Elimination System (NPDES) permit as provided in 40 CFR 125; NPDES permits may include limits for radionuclides (EPA, 1988b).In addition, individual states may have similar permitting requirements. The IAEA has described methods for calculating "release upper bounds" for effluents from uranium and thorium mines and mills (IAEA, 1989). These methods incorporate the use of analytical models and reflect limits on exposures of individuals residing near the facility.

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7. GUIDELINES,STANDARDS AND REGULATIONS

Wastes

NRC radiation protection regulations in 10 CFR 20 include radioactive waste disposal requirements (NRC, 1988);in addition, specific regulations for shallow land burial are contained in 10 CFR 61, "Licensing Requirements for Land Disposal of Radioactive Wastes" (NRC, 1988). Agreement states have comparable regulations. At the time of this writing, EPA was examining the need for further radiation regulations for wastes under the Solid Waste Disposal Act and the Resource Conservation and Recovery Act. 7.5.3

Uranium and Thorium Processing Sites

EPA has issued standards in 40 CFR 192 for both inactive and licensed commercial uranium and thorium processing sites (EPA, 1983a). The EPA National Emission Standard for 222Rnfrom Licensed Uranium Mill Tailings, 40 CFR 61, Subpart W (EPA, 1989a) requires work practices that control radon emissions and refers to the requirements of 40 CFR 192. NRC regulations for uranium mill tailings are contained in 10 CFR 40, Appendix A (NRC, 1988). 7.5.4

Other

The drinking water standards, 40 CFR 141adopted under the Safe Drinking Water Act (SDWA) and its amendments include limits for radionuclides (EPA, 1988~).The SDWA also requires that states establish underground waste water injection programs according to EPA regulations. The EPA has also established guidelines (EPA, 1990) for waste management practices for water treatment plant wastes containing radionuclides at concentrations in excess of average background levels.

8. Radiation Emergency Response Planning 8.1 General

Mining and milling facilities that have a need to control radioactive materials, also have a need to provide emergency planning and response capabilities for radiation accidents, after instituting efforts to eliminate or reduce the potential for such accidents [see NCRP Report No. 111(NCRP, 1991b)l. To mitigate the results from an accident, facility personnel should be trained to respond to emergency situations. The training should begin with the initial indoctrination on radiation and radioactive materials, and be complemented by routine updates and continuing education. The extent of emergency response and preparedness training appropriate to a specific facility should be commensurate with the potential consequences of a n accident. Emergency response plans should be documented and available to the staff, and the staff should be trained in the required procedures. One person must be clearly in charge of an emergency situation. Fire, police, security and medical personnel may be available from a nearby location (city, hospital, ambulance service, etc.). The responsibilities these personnel will have in emergency situations should be detailed and a radiation-safety training program devised to meet these responsibilities. This special training generally includes a description of the potential radiation exposure situations a t the facility and the possible effects associated with potential accidents. The intent of such training is to provide sufficient knowledge to permit informed judgments in emergency situations. Coordination with appropriate local, state and federal emergency response groups and radiation regulatory agencies is essential for an effective protective action. Some emergency response activities are required by federal, state andlor local government entities. 8.2 Operations

Due to the nature of the mineral extraction industry, many potential hazards exist a t such facilities. Some of these potential hazards

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8. RADIATION EMERGENCY RESPONSE PLANNING

contribute to a radiation accident situation. In addressing emergency response plans for radiation situations, provision should be made for decontamination of involved individuals that does not preclude medical lifesaving actions. Such a plan should assign responsibilities and establish notification procedures for both on- and off-site incidents. Facility planning efforts should be coordinated with those for emergency medical plans associated with transportation and treatment. NCRP Report No. 65 (NCRP, 1980b)contains specific information on the management of persons accidentally contaminated.

8.3 Environment

A planning effort will identify proper steps to be taken to mitigate the impact should a designed facility fail, endangering public health andlor the environment, regardless of the cause of the failure. Coordination with spill, fire and accident response authorities will help mitigate any impacts and hasten actions by other entities having further responsibilities. An emergency response plan should include impoundment management, environmental assessment, and preplanned cleanup procedures that would be used in addressing an accidental release. Notification of the NRC, the National Response Center, and other federal, state and local authorities may be required.

8.4 Transportation

While the contract carrier has the responsibility for the safe transport of the materials, the shipper has the expertise and interest in performing an appropriate clean-up operation in the case of a transportation incident. The use of consultants may be appropriate in addressing prevention, preplanning for addressing emergency situations and responding to an accident. Appropriate documents must be provided to the carriers to instruct them on preventive procedures (such as the bracing of the containers in the vehicle), necessary notification of authorities, and immediate protective actions until assistance arrives.

9. Radiation Protection in

Specific Applications

9.1 Introduction

The production of most minerals involves "conventional" mining methods, such as drilling, blasting, removing and hauling large quantities of ore material, and "conventional" milling methods such as crushing and grinding the ore, extracting the mineral of interest through concentration, precipitation or stripping, and packaging and shipping the concentrated product. In some operations, however, minerals are recovered through methods that do not involve all of these conventional steps. In still other operations, a specific mineral is recovered as a by-product of the principal mineral being extracted. Examples of these so-called "nonconventiona1"productionoperations include in situ solution mining, heap leaching and side-stream extraction. Several specific operations are addressed in this Section because radioactive minerals, uranium and thorium, are the product of interest or because the discussion addresses a radiation safety program which may be less complex than that for a typical uranium mine or mill. A scientific basis is being developed for the management of naturally occurring radioactive materials in oil and gas production. As described in Section 3.1 of this Report, concentrations of those materials may be sufficient to require the application of radiation protection controls a t oil and gas production facilities. Radiation protection a t such facilities is not further discussed in this Report. However,two examples of the emerging literature are by the American Petroleum Institute (API, 1992) and Wilson and Scott (1992),which give readers the opportunity to gain a perspective of the magnitude of radiation protection controls appropriate to some oil and gas operations. The radiation safety concepts and practices described below apply to the mining methods specifically stated but are also suggestive of those practices appropriate across the breadth of the mineral extraction industry.

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9.2 Heap-Leach Extraction

Heap-leach extraction is the application of liquid leaching agents to ores for the extraction of the mineral content. When the heap-leach process involves uranium, thorium or other radioactive materials, control of the radioactive materials in the groundwater should be evaluated. When the heap-leach extraction is conducted on the ground surface, a pad of relatively impervious material, such as several feet of compacted clay, often is constructed with a drain field on top and under the pad. Drain pipes are set in gravel to enhance the collection and drainage of liquids. The ore is placed on top of the prepared pad, and the leaching agent is applied by a sprinkler system to the top of the ore. The leachate is collected by the drain on top of the clay pad and is usually recycled until the mineral concentrations are sufficiently high to allow for efficient extraction. The drain under the clay pad is used to collect liquids that penetrate the clay pad. When a leach extraction process is conducted in mines, sprinkler systems are often installed to enhance the application of the leaching agent to the side walls and ceiling (ribs and back) of the mine. If sufficient pyrite exists in the ores to allow for the natural formation of the acidic liquid, the leaching liquid can be water. Otherwise, a water solution of sulfuric acid, other acids or bases is typically used for extraction. As the leaching operation is a wet process operable throughout most of the year, freezing weather excepted, little dust is created by the process. If dry ores are to be placed on the leach pad, water should be used to control dust during the placement of the ore on the pad. If the ores are moist, such precautions are not necessary. Throughout the heap-leach process, the application of leaching agent to the ore suppresses the generation of dust. Dust control may also be needed after leaching operations end or if they are interrupted. The potential for groundwater contamination exists for heap-leach operations. Due to the uncertainties associated with groundwater monitoring and potential groundwater users, a groundwater monitoring program is often advisable if only to demonstrate the lack of adverse effects on groundwater. As an example of a groundwater monitoring program applicable to the processing of ores containing radioactive materials, Table 9.1 presents a monitoring program for a uranium heap-leach extraction operation. The program serves only as an illustration because of the variation in the designs of extraction circuits and site-specific conditions such as the permeability of surface soils. The monitoring program is based on the analysis of parameters that will give the earliest warning that the groundwater under

9.3 INSM'UMIN'ERAL EXTRACTION

TABLE9.1-Illustmtion of a Type of sample Leachate on top of clay pad

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monitoring program for a uranium heap-leach facility. Sampling frequency Constituents analyzed

Semiannual for first year, then after major changes in extraction circuit

All chemicals and radionuclides

Liquids from drain under clay pad

Quarterly

Chemical and radionuclides with elevated concentrations in leachate

Groundwater from uppermost aquifer hydrologically down gradient

Annually

Mobile constituents observed in leachate, eg., U, Se, SO,, C1, As

Air

When dusting conditions prevail

Uranium series, eg., U, Th, Ra

the site may be impacted in the future. Once those warnings have been observed, the facility operators will have an opportunity to correct the problem before any serious and lasting effects on groundwater occur. The final product from a uranium heap-leach operation is a slurry which is shipped to a uranium mill for drying or shipped to a conversion facility. During storage and shipment, the yellowcake sluny will produce gamma radiation from the ingrowth of uranium progeny. Gamma monitoring of personnel may be appropriate. Air monitoring of heap-leach operations involving radioactive materials is not warranted unless there is a reasonable likelihood that dust will be generated.

9.3 In situ Mineral Extraction Underground ore bodies may be mined through in situ or solution mining. In this method, a leaching solution is injected through wells into the ore body to dissolve the mineral of interest. Production wells are pumped to bring the mineral-bearing solution to the surface where the mineral is extracted by appropriate methods such as ion exchange or solvent extraction. The solution is then treated for reinjection into the mining zone or is discharged to an evaporation pond. Since the late 1970's, in situ mining methods have accounted for an increasing percentage of uranium production. The use of this method generally presents less potential for radiation exposure and

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releases to the environment than in the conventional mining and milling method. The potential for underground exposure to airborne radon, radon progeny and ore dust is removed; the conventional grinding, crushing and other ore handling requirements are absent; only a small percentage [less than five percent (NRC, 1980)l of the radium is commonly brought to the surface by the solution; and the solid waste generated is quite small as no tailings are produced. Nevertheless, at some steps of the in situ operation, there is potential for exposure to airborne and external sources of radiation. Radon-222 gas entrained in the recovered solution and produced by decay of radium is available for release at locations such as solution storage and surge control ponds, surge tanks and exchange columns. If the extracted uranium is dried rather than kept in a slurry form, airborne particulates can be generated. Radium buildup on exchange columns can cause significant external exposure fields. The small amounts of solid wastes produced, such as materials filtered from the solution stream, pond and tank sediments, extraction and exchange column media and contaminated equipment, also contain radionuclides and are potential sources of exposure and environmental releases. The elements described in previous sections of this Report are all applicable to in situ uranium operations. Careful attention should be given to the layout of the facility to ensure isolation of and shielding of areas where elevated external exposure fields will be present. The equipment should be designed for ease of access, maintenance, removal and replacement, and should be constructed of materials that minimize buildup and collection of radioactive materials. Ventilation should be provided to prevent buildup of radon progeny. Operational practices should include access control, personnel protective equipment, special work permit requirements for designated activities and good housekeeping procedures. Training programs are an essential requirement. A groundwater monitoring program is necessary for in situ mining operations to identify the boundaries of the mining field, detect injection fluid that may have migrated from the leaching zone, and demonstrate restoration of the water in the mining zone at the end of the operation. The groundwater monitoring program should include analyses for radionuclides as a check on the integrity of the solution ponds used for storage of the solution and waste water from the process. 9.4 Side-Stream Extractions of Uranium

Side-stream extractions produce a mineral that is a by-product of the principal mineral being extracted. Even though uranium and

9.4 SIDE-STREAMEXTRACTIONS OF URANIUM

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its progeny may be present in low concentrations in the ore being milled, uranium, radium or other uranium progeny can concentrate in the side-stream extraction circuits and the product will be a concentrated form of uranium. The recovery of uranium from phosphoric acid is presented here as an example of side-stream extraction for which the technology is well developed, and the radiation protection considerations have been evaluated and defined. Other examples of side-stream uranium extraction processes include the extraction of uranium along with tungsten and tin from molybdenum or copper extraction operations. Usually, the potential occupational and environmental exposures from these circuits are low and do not by themselves warrant extensive protection programs and monitoring. However, sdiicient sampling should be conducted so that the operator knows what radionuclides are present in the circuit, the potential for exposuresand whether a radioactive materials license is required. Radiation protection program elements should be incorporated as appropriate based on this assessment. If the final product of the side-stream extraction is a slurry, the exposure potential will be similar to that from heap-leach operations. In contrast, if the final product is a dried powder, eg., yellowcake, then precautions and monitoring programs similar to those incurred in uranium mills should be considered (NRC,1983a; 1983b). 9.4.1

Uranium Recovery from Phosphoric Acid

The uranium associated with phosphate rock constitutes a significant potential source of this mineral. In the production of phosphoric acid by the wet process, the uranium contained in the input phosphate rock appears largely in the product acid. Thus, the mining, beneficiation and initial extraction of uranium have been performed in producing the phosphoric acid. The extent to which such facilities are operated depends upon the economics of the uranium industry; at least six different facilities were in operation in the United States at some time during the late 1970's and early 1980's. Uranium can be recovered from phosphoric acid by several variations of solvent extraction processes. In the multiple-extraction processes, several stages of organic extraction followed by aqueous stripping are employed to produce Ammonium Uranyl Tricarbonate (AUT). The AUT is filtered and calcined to a U,O, product which is drummed and shipped. In the single-cycle-extraction process, uranium is recovered by treating the phosphoric acid with a reducing agent, solvent

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extraction, purification of the organic phase, reacting with HF and precipitation and drying of UF, ("green salt"). The UF, may be drummed and shipped a t this stage. Alternatively, the UF, can be redissolved and ammoniated to produce Ammonium Diuranate (ADU) which is filtered and calcined to U308.The U308would then be drummed and shipped a s in the multiple-extraction processes. In the single-cycle-extraction (UF,) process, the product contains isotopes of thorium (primarily 234Thand 230Th)as well a s uranium. On the other hand, freshly precipitated AUT (multiple-extraction processes) and ADU (from further processing of UF,) contain uranium free of thorium and other decay products. In these processes, the radioactivity ofthe intermediates and the U30, product is essentially limited to uranium and early ingrowth products. The entire uranium recovery unit may be located a t the phosphoric acid plant. Alternatively, for the multiple-extraction process, the initial extraction and stripping may be located a t the phosphoric acid plant, with the first strip solution being transported to a central uranium recovery plant where the subsequent extraction, purification, production of U308, drumming and shipping take place. Facilities for recovery of uranium from phosphoric acid and the radiation protection considerations for such facilities are similar to those a t the extraction and subsequent stages of a conventional uranium mill. The major source of occupational exposure, as well as potential radioactive effluents, is dust generated in calcining and produd-packaging operations.

9.4.2

Occupational Exposure Considerations

External radiation levels are generally low in most of the uranium recovery facility. The maximum exposure rate is likely to be found in the product storage area where levels will depend upon the quantity and age of the product present. In addition, traces of 226Rathat occurred in the phosphoric acid may accumulate with time a t various locations in the process (such as in carbon purification columns, if used). Therefore, external radiation surveys should be performed on an occasional basis to identify any areas of elevated radiation levels, and personal monitoring should be provided for personnel regularly working in the facility. The primary source of occupational exposure is airborne dust a t the calcining and product drumming operations. In design of the plant, these operations should be isolated to reduce the area in which special restrictions may have to be employed for airborne radioactivity control and to facilitate control in the event of a n equipment

9.4 SIDE-STREAMEXTRACTIONS OF URANIUM

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malfunction. Drumming stations should be provided with exhaust ventilation for dust control. Radiation control zones should be designated on the basis of potential for airborne and surface contamination. A change room with "clean" and "controlled" sides and shower facilities should be provided a t the point of personnel access to the facility. Access control, protective clothing and special procedure requirements for contamination control zones should be specified. Respiratory protection should be provided for maintenance operations on dry product handling equipment, in the event of failure of dust control equipment, and at other times and locations indicated by airborne radionuclide levels. Routine air particulate monitoring, contamination monitoring and bioassay programs should be conducted following the guidelines of Section 5. Personnel should monitor themselves for contamination when leaving the contamination control area.

9.4.3

Shipping and Transportation

Where the uranium-loaded first-stage stripping solution is transported from the phosphoric acid plant to a central uranium recovery plant, emergency planning should include considerationsfor dealing with highway accidents. As indicated in Section 5, packages of dried product should be surveyed for surface contamination and decontaminated if necessary before shipment. Emergency planning should include planning for highway accidents of dried product shipments as indicated in Section 8.

9.4.4

Effluents and Environmental Monitoring

There should be no major liquid effluent from these facilities. The primary raffinate from solvent extraction operations is phosphoric acid depleted in uranium and this is returned to the phosphoric acid plant. However, some miscellaneous liquid waste streams (e.g., drains from process areas, changelshower rooms and laboratories) may contain radioactive materials. These should be identified and an appropriate hold-up, monitoring and treatment plan developed to control radionuclide discharges to the environment. As previously stated, essentially all of the radioactive material discharged from a uranium extraction plant is due to particulate matter that becomes airborne in t h e drying, calcination and

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9. RADIATION PROTECTION IN SPECIFIC APPLICATIONS

product-packaging areas. The drumming-station exhaust and other exhaust ventilation from the facility should be provided with an appropriate combination of filters and scrubbers to provide adequate control of airborne radionuclides. The environmental monitoring program should be designed to detect the effects of any airborne releases from the facility. The general design and conduct of the program should be consistent with the principles presented in Section 6. The primary radionuclide of interest is uranium. In the case of a facility utilizing the UF, process, 234Thand 230Thare also of interest.

9.4.5 Solid Radioactive Waste and Equipment Reuse or Salvage Solid radioactive wastes will consist of discarded process equipment (eg., filters, valves, miscellaneous hardware), protective clothing and chemicals (eg., solids from clarification stages) (Davis et al., 1979).Radioactive contamination of these wastes will be due primarily to natural uranium (and 234Th/230Th in the UF4 process) and early decay produds. Management of these wastes should be in conformance with the applicable state or federal regulations and would involve some combination of on-site retention andlor transfer to an off-site low-level radioactive waste facility. Process equipment should not be released for off-site reuse or salvage unless it has been thoroughly decontaminated and monitored to verify the efficacy of the decontamination.

9.5 Thorium and Rare-Earths Processing

Thorium occurs in significant concentrations (up to about eight percent) in mineral sands such as monazite (a brown or brownishred native phosphate of the cerium metals) and ilmenite (a lustrous black mineral, an oxide of iron and titanium). These sands can be processed for their thorium value but more typically are processed to recover other values such as rare earths and rutile (a mineral Ti02). The presence of thorium and thorium progeny indicate the potential for exposure to airborne radon and dust particulates and to external gamma and beta radiation fields. Further, since the market for thorium presently is small, most or all of the radionuclides initially in the sands are contained in the tailings. Thus, radiation management program considerations are appropriate for these operations and are comparable to those required for uranium extraction operations.

9.5 THORIUM AND RARE-EARTHS PROCESSING

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85

There are differences between the thorium and uranium decay series radionuclides that warrant particular attention. The thorium decay series is characterized by relatively short-lived progeny and the potential exposures from beta and gamma radiation are more pronounced than for the uranium series. For equal activity concentrations, there are about 30 percent more gamma emissions from the thorium series than from the uranium series and about 90 percent more energy is emitted. External exposure levels may range from less than 10 pGy per h to more than 100 pGy per h in monazite storage areas (IAEA, 1987). Af'ter processing, thorium and its radon progeny (220Rn)return to equilibrium much more rapidly than do uranium and its radon progeny (222Rn),SO the buildup of radon progeny from separated thorium presents a relatively greater exposure potential than from separated uranium. The initial processing steps for both thorium and rare-earths production are similar and include crushing and grinding, digestion, extraction of the mineral, packaging, storing and shipping and management of waste streams. The requirements for the radiation exposure management program are comparable to those for uranium processing activities and will vary in complexity and magnitude with the concentration of thorium in the feed material, the processes involved and the product produced. The radiation protection program should consider the basic features described in this Report. The layout of the process facility should take into account the external gamma and beta radiation sources and provide the conventional measures of isolation, limited access and shielding of areas where there is the potential for elevated exposure. These areas should include the material storage bins, grinding, crushing and product preparation locations, and waste locations. Equipment should be designed for ready access, maintenance, removal and replacement, and should be constructed of materials that retard accumulation of process constituents. Ventilation requirements should be evaluated and equipment installed for precluding buildup of radon and radon progeny and for controlling airborne dust. In this regard, consideration should be given to the potential for fragmentation of monazite during physical separation processes, resulting in the concentration of monazite in airborne dust. Operational practices should include the provisions described in Section 4 such as training of workers, access control,standard operating procedures and personal protective equipment. Appropriate procedures for radiation monitoring and surveying should be developed and should recognize that different counting intervals are required

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9. RADIATION PROTECTION IN SPECIFIC APPLICATIONS

to distinguish "'Rn and "'Fin progeny levels (Section 5). Further, bioassay for radionuclides in urine is not likely to be informative, and the use of more sophisticated techniques will be needed for the assessment of exposures. The wastes, especially the tailings, should be isolated from the environment to preclude unnecessary exposure of workers and the public. Other wastes, such as worn out and replaced equipment, filters, piping and debris should also be surveyed for the presence of contamination and decontaminated to accepted release levels or disposed of in accordance with acceptable practices.

9.6

Phosphate

Uranium and its progeny are associated with phosphate deposits of marine origin. Consequently, radiation protection requirements should be considered for phosphate mining and processing operations. In the United States, commercial phosphate mining and beneficiation take place primarily in Florida with additional operations occurring in states such as North Carolina, Tennessee, Idaho, Montana, Wyoming and Utah (Owen and Hyder, 1980; EPA, 1984a). The extracted mineral, phosphate rock, is further processed in chemical plants a t a variety of locations around the country to produce phosphoric acid, fertilizer ingredients, animal feed ingredients, elemental phosphorus and other phosphate products. Radiation assessment is facilitated by dividing the industry into five segments: (1) mining, beneficiation and wet rock handling, (2) phosphate rock drying and dry rock handling, (3) wet process phosphoric acid production, (4) production of phosphate products, and (5) elemental phosphorus production by the thermal process.

9-6.1 Mining, Beneficiation and Wet Rock Handling Phosphate deposits occur at varying depths. Present operations in Florida consist of surface mining of deposits with overburden thicknesses to 15 m. The overburden is cast aside by draglines to expose the matrix (the useable phosphate ore)which is then removed. In the undisturbed state, the uranium decay series is essentially in radioactive equilibrium in the overburden and the phosphate deposit. In a typical Florida profile, concentrations of series members

9.6 PHOSPHATE

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87

are low at the surface, on the order of 20 to 40 Bq per kg, and increase gradually with depth to values on the order of 700 to 4,000 Bq per kg immediately above or in the matrix. In Florida and North Carolina, the matrix is beneficiated to increase the phosphate content of the commercial product, which is termedphosphate rock. On the other hand, almost half of the western phosphate ore is used directly without any beneficiation (Owen and Hyder, 1980). Commercial phosphate rock, from various origins in the United States, contains on the order of 150 to 4,800 Bq per kg of 238U and on the order of 10 to 80 Bq per kg of 232Th(Menzel, 1968; Guimond and Windham, 1975; Roessler et al., 1979a; Owen and Hyder, 1980; UNSCEAR, 1982). When beneficiation is performed, washing, screening and flotation separation produce several sizes and grades of marketable phosphate rock product, a clay suspension waste and a sand tailings waste. The sand tailings are returned to the land while the phosphatic clay suspension is pumped to clay settling areas. In central Florida, dryweight concentrations of the individual members of the uranium decay series are on the order of 70 to 400 Bq per kg in sand tailings and 550 to 750 Bq per kg in waste clays. The beneficiated phosphate rock product ("wet rock") is stockpiled in large outdoor piles, usually near the plant. The product is removed from a pile for sale or for transfer to a phosphate chemical plant via a conveyor in a wet rock loading tunnel which may be at ground level or totally or partially below ground.

9.6.1.1 Occupational Exposure. The major occupational radiation exposure concern in this portion of the industry is the potential for airborne radon and progeny in rock-loading tunnels and other enclosed spaces containing significant quantities of phosphate rock (Roessler and Prince, 1978; Roessler, 1981). Other candidate areas for elevated radon concentrations include rock storage bins or silos and compartments in rock-shipping vessels. Radon progeny concentrations usually will be low in work areas that are open to the atmosphere or well ventilated. Adequate natural or mechanical ventilation should be provided to control airborne radon levels in rock-loading tunnels. Other enclosed spaces containing high inventories of phosphate rock should be ventilated prior to entry of personnel. An initial radon or radon progeny survey should be performed in tunnels and other potential locations of airborne radon to determine the range of levels in the facility. Additional surveys should be performed:

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9. RADIATION PROTECTION IN SPECIFIC APPLICATIONS

(1)periodically (every one to two years) to verify that acceptable conditions are being maintained, (2) following any design or operating changes that have the potential for increasing the radon source term and (3) following any modifications intended t o reduce radon concentrations. In mining and beneficiation, gamma radiation !evels range from a normal background of 50 to 100 nGy per h over unmined land and in the dragline cab to 1 pGy per h in the near vicinity of large quantities of beneficiated rock. This exposure route is not a major consideration, annual effective dose or equivalent dose will not exceed 1mSv per y above background under typical use and occupancy conditions. A walk-through gamma radiation survey provides assurance that levels are as described above, however, follow-up surveys are optional. Other potential radiation sources include industrial gauges containing sealed radioactive sources as described in Section 4. Mining and wet rock operations are not particularly dusty and airborne long-lived radionuclides are not likely to be of concern. If obviously dusty operations are observed, air samples should be collected and analyzed for long-lived radionuclides and a n assessment made of the inhalation exposure by personnel and the need for dust-abatement procedures. 9.6.1.2 Mining and Beneficiation Wastes and Post-Mining Land. Surface mining, disposal of beneficiation waste materials and land reclamation involving various combinations of overburden spoils, sand tailings and settled clays have the potential for producing waste material stockpiles and near-surface land forms with elevated levels of radioactivity. While these materials do not exhibit direct radiation levels which warrant substantive radiological controls, they can be a source of radon. The surface radioactivity levels of mined lands should be minimized to the extent practical in the mining process by placement of the deeper, higher radioactivity spoils as deep as possible with a cover of lower radioactivity material. Mining spoils and beneficiation wastes with elevated concentrations of 226Rashould not be used as fill for the construction of buildings. When mined lands are used for building construction, construction designs should be employed that minimize the entry of radon from the soil (NCRP, 1984a). NCRP Report No. 103(NCRP, 1989c)recommends techniques for increasing the resistance of building fabric to soil gas entry. A general term for these techniques is sealing.

9.6 PHOSPHATE

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89

State and local requirements for pre- and post-mining land monitoring and regulations regarding the use of mined lands and of mining and beneficiation wastes should be consulted.

Liquid Releases and Water Quality. Concentrations of dissolved radionuclides in water from mining and beneficiation would be expected to be low. However, these waters, especially from phosphatic clay slurries, may contain suspended matter with elevated levels of radioactivity (Guimondand Windham, 1975).Effluent problems are minimized if these waters are recycled through the processes to the greatest extent possible. Any water that is to be released should be provided with sufficient settling and clarification to remove the radionuclide-bearing suspended matter. Waters that are released should be monitored for radionuclides and the monitoring results compared to the nonoccupational dose limits and the requirements of applicable permits and water quality standards. 9.6.1.3

9.6.2 Phosphate Rock Drying and Dry Rock Handling

Although there is an increasing use of wet rock as feed for chemical processing, much of the wet rock production is dried and ground for shipment andlor processing. 9.6.2.1 Occupational Exposure.

In the vicinity of dryers, grinding mills and dry rock loading and unloading areas, there is a potential for airborne long-lived radionuclides to be associated with the dust (Windham et al., 1976;Roessler, 1981).These areas should be evaluated on a case-by-case basis, and dust control measures should be employed where needed. Air sampling should be conducted in these areas following any design or operating change likely to increase airborne dust, following any modifications intended to reduce dust levels, and on a periodic basis to verify that acceptable conditions are being maintained. For areas showing significant gross-alpha concentrations, specific radionuclide analyses can aid in interpretation of the radioactivity present. Personal respiratory protection should be provided for short-term tasks involving temporary, unusual dust concentrations.

Emissions. In a study of a large Florida rock drying facility, plant-attributable levels of radioactivity in air surrounding the plant were estimated to produce individual lung doses of less than 0.1 mSv per y for persons living in the immediate area (Partridge et al., 1978). These observations suggest that periodic monitoring of the radioactivity of emissions from dry rock facilities is not

9.6.2.2

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9. RADIATION PROTECTION IN SPECIFIC APPLICATIONS

warranted. Maintenance of the effluent a t levels which are as low as reasonably achievable should be evaluated and appropriate actions taken. 9.6.3 Wet-Process Phosphoric Acid Plants

Much of the phosphate rock production is further processed in web process chemical plants to produce phosphoric acid. The phosphate rock (a calcium phosphate mineral) is reacted with sulfuric acid and the resulting slurry is then filtered to separate the product phosphoric acid from the insoluble by-product phosphogypsum (2CaS0, H20or CaSO, H20).The filtered phosphogypsum is reslurried and pumped to a storage area where the acidic water is allowed to drain away into a pond, and the drained phosphogypsum accumulates in a pile. In phosphoric acid production, the radioactive equilibrium existing in the phosphate rock is disrupted, with the uranium and thorium appearing primarily in the phosphoric acid and the radium, 'lOPb and 210Potending to be found with the phosphogypsum. Radon-226 can be concentrated in scales and sediments in various places in the digestion, filtration, cooling and acid-receiving systems of wet-process phosphoric acid plants (Lardinoye et al., 1982; Windham et al., 1976; Roessler et al., 1979b; Keaton, 1987). Levels observed in Florida were quite variable and ranged from a few hundreds of Bq per kg in acid reaction vessels and phosphoric acid tanks to hundreds of kBq per kg in filter pans. Radiological considerationsassociatedwith phosphoric acid plants include: (1) potential occupational exposure during regular operations, (2) occupational exposure during maintenance and cleaning, (3) occupational exposure during filter pan refurbishing, (4) management of plant and equipment clean-up and maintenance wastes and ( 5 ) environmental considerations associated with management and possible use of the by-product phosphogypsum. 9.6.3.1 Occupational Exposure-Pmtection Operations. The principal radiation exposure in a phosphoric acid plant is related to the buildup of in the residues. Elevated gamma radiation levels have been found in some Florida facilities with calculated radiation doses up to 0.4 mSv per week (Windhamet al., 1976;Roessler, 1981). In keeping with the principle of ALARA, the design of new plants should locate control panels and other highly occupied work stations away from filter pans and filtrate receiving tanks.

1

91 In operating plants, external gamma radiation surveys should be performed on a periodic basis (every one to two years) to verify that acceptable conditions are being maintained and to provide a current knowledge of conditions in the plant. Any airborne radon problems in a phosphoric acid chemical plant would be expected to be limited to those already described for entry of enclosed unventilated spaces with significant inventories of phosphate rock. Any long-lived airborne radionuclide problems would be those of dusty dry rock handling operations as described previously. 9.6 PHOSPHATE

9.6.3.2 Occupational Exposure--Clean-up and Maintenance. Maintenance and clean-up operations involve work in the vicinity of radium-bearing residues during such operations as filter cloth replacement, filter pan repair, scale removal in the reaction vessels (attack tanks) and replacement of piping and other system components. Surveys of radiation levels experienced by maintenance personnel should be performed periodically. Individual assignments of special tasks involving close contact to residues in attack tanks, fiIters, piping and filtrate receiving tanks should be evaluated by considering gamma exposure rates and anticipated occupancy times to estimate likely cumulative exposures and to determine whether personnel monitoring or regular surveys are indicated. Such work should be preplanned to minimize occupational exposure.

Occupational Exposure-Filter Pan Repair. The greatest potential for exposure to radiation and airborne radionuclides occurs in filter pan refurbishing, either a t the plant site or a t off-site machine shops. Equipment is taken out of service, internal parts with radium-bearing scale are removed from the inherent shielding of the system, and drying removes the shielding and airborne-radionuclide-suppression effects of the liquid. External gamma radiation levels that might be expected in filter pan cleaning and maintenance range from 10 pGy per h in the general vicinity of filter pans to 120 pGy per h at contact with an uncleaned pan (Keaton, 1987). Cumulative radiation doses to personnel would depend upon the type, number and spacingof pieces of equipment in the work area, the contributionfrom general contamination and the time spent in the various radiation fields. Work in close proximity to filter pans for 2,000 h per y would result in annual equivalent dose intermediate between the 1mSv per y recommendation for individuals of the general public and the 50 mSv per y limit for radiation workers.

9.6.3.3

,

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9. RADIATION PROTECTION IN SPECIFIC APPLICATIONS

There is also the potential for elevated levels of airborne radionuclides in the repair facility as a result of procedures such as cleaning, cutting and grinding. Filter pans and other equipment removed from service in a phosphoric acid plant should be surveyed for gamma radiation as an indication of the relative quantities of radium contamination. If elevated levels of radioactivity are detected, a RSO should be appointed and a radiation control and monitoring program established. Criteria for "elevated" should be developed on the basis of experience. A suggested criterion is a near-contact gamma radiation level twice the general plant background when measured with a gamma scintillation or sensitive GM survey meter. 9.6.3.4 Waste Management. Solid wastes potentially containing radioactive contamination include: (1) residues and scales from cleaning of plant equipment; (2) disposable items such as filter cloths and miscellaneous metal, rubber and other scrap; and (3) piping and equipment that have been removed from service. Such materials should be surveyed to determine the relative degree of contamination. Low-level contaminated bulk residues should be transferred to the phosphogypsum pile. Residues and scrap with higher levels of contamination present a different issue. One practical solution is to contain these materials a t the chemical plant. In the case of off-site repair shops, this would mean transferring wastes back to the plants a t which the contaminated equipment originated. (Transportation regulations and recommendations should be reviewed in planning such work activities.) Options include storage a t a designated place on the gypsum pile or storage in containers a t some other designated location on the plant site. Equipment and materials that have been taken out of service in phosphoric acid chemical plants should not be released to the general salvage flow stream. Radiation surveys should be performed and the items should be released only if not contaminated. If contamination is detected, the item should be decontaminated or handled as contaminated waste and prevented from entering the salvage flow stream. Liquid wastes potentially bearing radioactive materials should not be released directly to noncontrolled areas. Liquids from cleaning operations at the plant can be co-mingled with the gypsum slurry and gypsum pond water which is then either contained and recycled at the plant or is treated to remove radionuclides before release. Liquid wastes from off-site shops which contain elevated levels of radionuclides must be treated to remove the radionuclides before

9.6 PHOSPHATE

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release. Sediments and sludges from water treatment should then be treated in a similar fashion to the solid radioactive residues. 9.6.3.5 Phosphogypsum. A large phosphoric acid plant annually produces on the order of 1.2 x lo6 m3 of phosphogypsum (EPA, 1979).Phosphogypsum derived from central Florida rock has a 226Ra concentration on the order of 0.7 to 1kBq per kg. Large quantities of this material, such as in storage piles, will have gamma radiation levels producing an equivalent dose rate on the order of 300 nSv per h at one m from the surface. Similar data have been reported for an operation in Louisiana (Laiche and Scott, 1991).The drained areas of active storage piles will have a radon exhalation rate on the order of 0.7 to 1Bq m"s-l; this eventually decreases to less than 0.4 Bq m-2s'1 for inactive storage piles. Radiation and radon flux values associated with phosphogypsum derived from other sources would be roughly proportional to the 226Raconcentration of the input phosphate rock. Airborne radon concentrations on the order of 10 to 30 Bq m*3 (5 to 25 Bq m-3 above background) were observed over and in the immediate vicinity of Central Florida piles (Roessler, 1987; Horton et al., 1988). Airborne radon progeny concentrations of less than or equal to 0.001 WL have been reported at active piles in Florida and Louisiana (Roessler, 1987; Laiche and Scott, 1991). Personnel working on phosphogypsum piles for 2,000 h per y would have annual gamma radiation doses not exceeding 0.7 mGy per y; annual doses would be proportionally lower for lower occupancy times and would be considerably lower for work with small amounts. The water from the phosphogypsum slurry is highly acidic and contains 226Raconcentrations on the order of 2 to 4 kBq per m3 (Guimond and Windham, 1975;NFIC, 1973).Ordinarily, pond water is recycled for plant use. If it is necessary to release water, such as in the case of excess quantities due to rainfall, it should be treated to remove the radium and the release should be sampled and analyzed for radionuclide content. Liming to control the acidity is also effective in removing the dissolved radium (Guimond and Windham, 1975). Phosphogypsum is a potential raw or resource material for productive uses. Proposed uses as building material should be preceded by an assessment of the potential contribution to indoor radon and gamma radiation exposure. Other uses include as an aggregate or an aggregate ingredient in nondomiciliary construction and for application to agricultural land as a mineral source or as a soil amendment. On the basis of limited assessments to date, these latter uses do not appear to present major routes for radiation exposure

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9. RADIATION PROTECTION IN SPECIFIC APPLICATIONS

(TDH, 1983; Lindekin and Coles, 1978; Lindekin, 1981; Roessler, 1986). However, such uses should be accompanied by evaluation of the potential radiation exposure. 9.6.4

Production of Phosphate Products

Phosphate rock and phosphoric acid are used as the raw materials for such products as fertilizer and animal feed ingredients. In these operations, minimal gamma radiation exposures would be expected downstream from the phosphoric acid plant. This can be verified by a gamma radiation survey. If feed materials or products with elevated 226 Ra concentrations are held in enclosed, poorly ventilated spaces, airborne radon monitoring should be performed and, if indicated by the monitoring results, ventilation provided. Long-lived airborne radionuclides may occur at dry rock input to the process, a t product grinding and sizing, or in product handling, storage and loading. If exposure conditions exist, dust control measures should be applied and periodic air sampling performed following guidelines outlined earlier in this Report.

9.6.5

Thermal Process (Elemental Phosphorus)

In elemental phosphorus production by the thermal process, phosphate rock, coke and silica are treated in an electric furnace to liberate phosphorus as a vapor which is condensed to liquid phosphorus. Two by-products are slag (primarily calcium silicate) and ferrophosphorus (or "ferro-phos"), a mixture of iron phosphides. In many thermal process operations, the phosphate rock is calcined at high temperatures and briquetted to provide a better feed material for the furnace. The calcining and the furnace volatilize a significant fraction of the 'lOPb and zloPowhich appear as deposits in stack linings and as airborne emissions (EPA, 1984a; 1984b).The balance of the radionuclides from the input rock appear in the slag and ferrophos. Occupational exposure studies in thermal process facilities have been reported for Florida (Windham et al., 1976;Roessler and Prince, 1978; Roessler, 1981)and Idaho (EPA, 1977).External gamma radiation levels corresponded to 0.1 to l p,Gy per h and maximum annual whole-body effedive doses were estimated to be less than 3 mSv per y. In the Florida study, all observed airborne radionuclide concentrations were a t least an order of magnitude below occupational concentration limits. In the Idaho study, it was recommended that radon

9.6 PHOSPHATE

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andlor radon progeny measurements be made to better define exposure to personnel in the plant. Potential exposure levels should be assessed and appropriate protective measures defined before maintenance work commences inside furnaces and stack piping. Potential airborne emissions from calcining and furnace operations should be evaluated and, if necessary, appropriate control and monitoring instituted.'' Electric furnace slag from central Florida and similar rock contains 226Raconcentrations on the order of 2 kBq per kg. This material has been used as an aggregate in building products, road-building and septic tank drain fields and as a crushed rock or "ballast" for railroad beds and electrical switching yards. This material will produce higher direct radiation levels than most aggregate and building materials. On the other hand, slag has a low emanating coefficient, and radon emanations from this material will be much lower than those associated with other materials with the same 2xRa concentrations (Roessleret al., 1979b;Lloyd, 1983). Proposed uses ofphosphate furnace slag should be evaluated to ensure that any associated radiation exposures can be justified.

''The EPA has evaluatedpotential airborne emissionsfrom thermal proceas plants (EPA, 1984a; 198415) and published standards pertaining to emission limits in 40 CFR 61 (EPA, 1985; 1989a).

APPENDIX A

Radioactive Series: 238U, 232Thand 235U There are three naturally occurringradioactive decay series which may be found in ores. They are denominated as the uranium series, the thorium series and the actinium series. These series are headed, respectively, by the very long-lived 238U,232Thand 235U.Where natural uranium ores are found, 0.7 percent of the uranium will be 235U. Consequently, members of two series will be found together, however, members of the 238Useries will predominate. Indeed, in most rocks and soils, members of all three series will be found with their relative concentrations dependent on physical and chemical processes ongoing in the environment. Figures A.l, A.2 and A.3 describe the radioactive series. Each begins with the long-lived parent, contains an isotope of the noble gas radon and ends with a stable nuclide of lead. The radon isotopes are the most mobile because of their unreactive, inert gas structure.

1

RADIOACTIVE SERIES

97

4.8 MeV

1: I 5.5 MeV

21W0 (%A) 3.05 min 6.0 MeV Alpha Decay

1,

214Pb(R8B) 26.8 min 0.7. 1.0 MeV

(wc')

z190(RaF)

214~0

138 d ; ~ 5 . 3MeV

1 . 6 ~ 1 0 "s i1.7 MeV

,

*148i ( ~ a c j 19.7 rnin 0.4-3.3 MeV

1

Z'oBi

2lQPb(RaD) 22 Y ~ 0 . 1MeV

Fig. kl. Principal decay scheme of the uranium series.

(RaE)

5.0d ;y1.2 MeV

- .-

1

206Pb(RgG) Stable

/

APPENDIXA

4.0 MeV

'7-kzd 1<0.i MeV

I

1.91 y 5.3,5.4 MeV

I 224Ra (ThX) 3.64 d 5.7 MeV

220Rn (Tn) 6.3 MeV

2 1 6 P(ThA) ~ 0.15 s 6.8 MeV

I Beta Decay

>,

212Po (ThC')

3x1W7s

213 11 8.8 MeV 212Bi(ThC) ' 60.6 rnin 2.2 MeV P 6.1 MeV a

2I2Pb ( ~ h ~ f 10.6 h 0.3, 0.6 MeV

1

1 ,I# 113

208Tl(The") 3.1 min 1 .O-1.8 MeV

Fig. A.2. Principal decay scheme of the thorium series.

msPb (ThD) Stable

RADIOACTIVE SERIES

171x108yl 4.4 MeV

231 Pa

3.2 x1 O4 y 5.0 MeV

25.5 h

18.2d

1 5.55.7 MeV I

6.4-6.8 MeV

2 l lBi (AcC)

Beta Decay Alpha Decay

lPb (AcB) Stable

1.4, 0.5 MeV 4.79 min 1.44 MeV

Fig. A.3. Principal decay scheme of the actinium series.

1

99

APPENDIX B

Conversion Factors TABLEB.I-Conversion factors between Sl and mnventwnal units (NCRP, 19853). Special name for SI unit becquerel

F~Y

s~evert

Symbol using special name

Conventional unit

%

curie rad rem

CY Sv

symbol for conventional unit Ci rad rem

1mCi = 37 MBq 1 pCi = 37 kBq 1nCi = 37 Bq 1pCi = 0.037 Bq

1 Bq = 1disintegration per s 1 Ci = 3.7 x 10" disintegrations per s

1 rad = 10 mGy 1mrad = 10 pGy

100 mrad = 1 mGy; 1 Gy = 100 rad 100 prad = 1 pGy

1rem = 10 mSv 1 mrem = 10 pSv

100 mrem = 1 mSv; 1 Sv = 100 rem 100 prem = 1 $ 3 ~

Value of conventional unit in SI units 3.7 x 10" Bq 0.01 Cy 0.01 Sv

SI prefures Factor 10l8 10l6 10'~

lo9 lo6 lo3 1o2 10'

Prefix exa

Symbol

E P

tera

T

gigs

G M k h

mega kilo hecto deka

da

Factor

lo-' lo-2 lo-3 lo4 lo-* 10-l2 10-l6 10-Is

Prefix deci centi milli micro nano pic0 femto atto

Symbol d c m CL

n P f

a

Glossary absorbed dose: The energy from ionizing radiation absorbed per unit mass is called the absorbed dose. The unit of absorbed dose is the gray (1Joule per kg) or, historically, the rad which is equal numerically to lo-' kg-' (100 erg 5'). air kerma: The sum of the initial kinetic energies of all the charged ionizing particles liberated by uncharged ionizing particles per unit mass of a specified material. Kerma is measured in the same unit a s absorbed dose. The SI unit of kerma is joule per kg and its special name is gray (Gy). Kerma can be quoted for any specific material a t a point in free space or in any absorbing medium. ALARA: An acronym referring to one element of the system of dose limitation, i.e., the recommendation to keep all radiation exposures as low as reasonably achievable, economic and social factors being taken into account. annual reference levels of intake (ARLI):The activity of a radionuclide that, taken into the body during a year, would provide a committed effective dose to a person, represented by Reference Man, equal to 20 mSv. The ARLI is expressed in becquerel (Bq). becquerel (Bq):A unit of radioactivity. One becquerel is one nuclear transformation per s. beneficiation: Preliminary conditioning of an ore for refinement. beta decay: Radioactive decay in which a beta particle is emitted or in which orbital electron capture occurs. bioassay procedure: A procedure used to determine the kind, quantity, location andlor retention of radionuclides in the body by direct (in uiuo) measurements or by in vitro analysis of material excreted or removed from the body. committed effective dose: Following a n intake to the body of a radioactive material, there is a period during which the material gives rise to equivalent doses in the tissues of the body a t varying rates. The summation of these committed equivdent doses is the committed effective dose. contamination: Deposition of radioactive material in any place where it may make products or equipment unsuitable for some specific use. The presence of unwanted radioactive material. curie (Ci):A special unit of radioactivity in the conventional system. One curie equals 3.7 x 101° nuclear transformations per s.

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GLOSSARY

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 DRAC is Bq per m3. The DRAC corresponds to the formerly used term maximum permissible concentration (MPC). dose rate: The radiation dose delivered per unit time. dosimetry:The measurement or calculation of the energy absorbed by matter. effective dose: The sum over all exposed tissues of the products of the equivalent dose in a tissue and the weighting factor for that tissue (NCRP, 1993). environmental exposure: Exposure to radiation through environmental pathways. equilibrium: A state in which the activity of all the progeny within a decay series is equal to the parent activity. For radon progeny, equilibrium is rarely achieved and the progeny activities are usually less than the radon activity. equivalent dose: A quantity used for radiation-protection purposes that takes into account the different probability of effects which occur with the same absorbed dose delivered by radiations with different radiation weighting factors. It is defined as the product of the average absorbed dose in a specified organ or tissue and the radiation weighting factor. The unit of equivalent dose is joules per kg and its special name is the sievert (Sv). exposure: In this Report, exposure is often used in its more general sense and not as the specifically defined radiation quantity. In the formal sense, a measure of the quantity of x or gamma radiation based on its ability to ionize air through which it passes. gamma ray: Electromagnetic radiation frequently accompanying alpha and beta emissions as radioactive materials decay. grab samples: Samples of limited volume taken at random or at preselected frequencies. gray (Gy):The unit of absorbed dose. 1 Gy = 1Joule per kg. half-life: The time interval during which one-half of the initial number of atoms of a radioactive nuclide undergo radioactive decay. After two half-lives, one-fourth of the initial atoms remain undecayed, after three half-lives, one-eighth remain, and so on. heap-leach extraction: The application of chemical agents to ore stockpiles or mine walls for the extraction of the mineral content. indirect bioassay: The assessment of radioactive material deposited in the body by detection of radioactivity in material excreted or removed from the body (in vitro measurement). intake: The amount of radioactive material taken into the body by inhalation, absorption through the skin, ingestion or through wounds.

GLOSSARY

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leach: Dissolve soluble material by the action of percolating liquid. leachate: The solution or product obtained by leaching. maximum permissible concentration (MPC): The concentration of a radionuclide in air or water that would lead to an amount of the radionuclide in the critical organ that would just deliver the maximum permissible dose rate to that organ. A phrase used when the conventional system of units is employed. It corresponds to the Derived Reference Air Concentration (DRAC). monitoring: Continuous or periodic determination of the amount of radiation or radioactivity present in a given area or in a volume of effluent. National Response Center: The federal government center to which releases of hazardous substances should be reported. The Center is operated by the U.S.Coast Guard and provides technical assistance and communications support for responding federal agencies. natural background radiation:The radiation in the earth's natural environment, including radiation originating outside the earth's atmosphere and radiation from the naturally occurring radioactive elements on earth. These elements may be found both in the environment and inside the bodies of men and animals. nuclide: A species of atom characterized by the constitution of its nucleus. occupational exposure: Exposure of the worker that is directly attributable to the worker's occupation. phosphate rock: An ore from which phosphorus is extracted and which often contains low concentrations of uranium. rad: A unit of absorbed dose in the conventional system. In the SI system of units 1 Gy = 1J kg-' (see absorbed dose). radiation: The emission and propagation of energy through matter or space. radiation safety (radiation protection):Concerned with recognition, evaluation and control of risks due to radiation exposure. radioactive: Exhibiting radioactivity or pertaining to radioactivity. radioactive decay: The spontaneous transformation of one nuclide into a different nuclide or into a different energy state of the same nuclide. The process results in a decrease, with time, of the number of the radioactive atoms in a sample. Decay generally involves the emission from the nucleus of alpha particles, beta particles or gamma rays. radioactive series: A succession of nuclides, each of which transforms by radioactive decay into the next until a stable nuclide results. The first member is called the parent and the subsequent members are called progeny, daughters or decay products.

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radioactivity:The spontaneous decay on disintegration of a n unstable nucleus, usually accompanied by the emission of alpha particles, beta particles or gamma rays. radionuclide: A nuclide that is radioactive. radon: A radioactive noble gas; in this Report, either 222Rnor 220Rn. radon flux: The number of radon atoms passing through a unit cross-sectional area per unit time. radon progeny (radon daughters):The short-lived radionuclides formed as a result of decay of radon. For 222Rn,they consist of '18Po (RaA), '14Pb (RaB), '14Bi (RaC) and '14Po (RaC'). Combined numbers of these radionuclides are reduced by one-half approximately every 30 min. raffinate: Fluid from the purification step in a mill, depleted in the mineral of interest relative to the fluid entering the purification process. The raffinate may be reused in the process stream or discarded. reference man: A person with the anatomical and physiological characteristics defined in the report of the ICRP Task Group on Reference Man [ICRP Publication 23 (ICRP, 197511. risk: The product of probability of occurrence and severity of injury, damage or loss. sampling: The process of taking a representative small portion or quantity of something for testing or analysis. side-stream extraction: The extraction of a mineral that is a byproduct of the principal mineral being extracted. sievert (Sv): The unit of equivalent dose. tailings: Waste or refuse left i n various processes of milling or mining. Tailings often contain a major portion of the radioactive materials present in the undisturbed ore. thermoluminescent dosimeter (TLD): A dosimeter using a material which upon heating emits light in direct proportion to the amount of radiation to which the material was exposed. thorium: A naturally radioactive element. Thorium-232 is the parent of one radioactive series, and specific thorium nuclides are members of the three radionuclide series (see Appendix A). uptake: Quantity of a radionuclide taken up by the systemic circulation, e g . , by absorption from compartments in the respiratory or gastrointestinal tracts. uranium: A naturally radioactive element. In natural ores, it consists of 0.7 percent 235U,99.3 percent 238U,and a small amount of 2 3 4 u

working level (WL):A unit of air concentration of potential alpha energy released from radon and its short-lived progeny. One working level is any combination of short-lived radon daughter

GLOSSARY

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products in one liter of air that will result in the emission of 1.3 x lo5 MeV (million electron volts) of potential alpha energy. 1WL will result in the emission of 2.08 x joules of energy. yellowcake: A product of uranium mills, concentrated in uranium content and suitable for shipment for further processing into fuel for reactors. The product is often in the form of a uranium oxide

(u,o,).

yellowcake drumming: The process of putting dried yellowcake into containers (drums) for storage and shipment.

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LLOYD, L.L. (1983). Evaluution of Radon Sources and Phosphate Slag in Butte, Montana (Montana Department of Health and Environmental Services, Butte, Montana). MENZEL, R.G. (1968). "Uranium, radium and thorium content in phosphate rocks and their possible radiation hazard," J. Agri. Food Chem. 16, 231-234. MILLS, W.A., FLACK, D.S., ARSENAULT, F.J. and CONTI, E.F. (1989). "A compendium of major U.S. radiation protection standards and guides: Legal and technical facts," ORAU 89/F-111(Oak Ridge Associated Universities, Oak Ridge, Tennessee). MSHA (1980). Mine Safety and Health Administration, U.S. Department of Labor. "Health and safety standards-underground metal and nonmetal mines" in Mineml Resources, 30 CFR 57 (US. Government Printing Office, Washington). MSHA (1986). Mine Safety and Health Administration, U.S. Department of Labor. Ionizing Radiation Standards for Metal a n d Nonmetal Mines; Proposed Rule, Federal Register 51, No. 244, 45678-45688 (U.S. Government Printing Office, Washington). NCRP (1976). National Council on Radiation Protection and Measurements. Environmental Radiation Measurements, NCRP Report No. 50 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). 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 a n d Monitoring Methods for Radiation Protection, NCRP Report No. 57 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1980a). National Council on Radiation Protection and Measurements. Perceptions of Risk, NCRP Proceedings No. 15 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1980b). 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 (1983). National Council on Radiation Protection a n d Measurements. Operational Radiation Safety-Training, NCRP Report No. 71 (National Council on Radiation Protection and Measurements, Bethesda, Maryland). NCRP (1984a). National Council on Radiation Protection and Measurements. Exposures from the Uranium Series with Emphasis on

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The NCRP

The National Council on Radiation Protection and Measurements is a nonprofit corporation chartered by Congress in 1964 to: 1. Collect, analyze, develop and disseminate in the public interest information and recommendations about (a) protection against radiation and (b) radiation measurements, quantities and units, particularly those concerned with radiation protection. 2. Provide a means by which organizations concerned with the scientific and related aspects of radiation protection and of radiation quantities, units and measurements may cooperate for effective utilization of their combined resources, and to stimulate the work of such organizations. 3. Develop basic concepts about radiation quantities, units and measurements, about the application of these concepts, and about radiation protection. 4. Cooperate with the International Commission on Radiological Protection, the International Commission on Radiation Units and Measurements, and other national and international organizations, governmental and private, concerned with radiation quantities, units and measurements and with radiation protection.

The Council is the successor to the unincorporated association of scientists known as the National Committee on Radiation Protection and Measurements and was formed to carry on the work begun by the Committee. The Council is made up of the members and the participants who serve on the scientific committees of the Council. The Council members who are selected solely on the basis of their scientific expertise are drawn from public and private universities, medical centers, national and private laboratories and industry. The scientific committees, composed of experts having detailed knowledge and competence in the particular area of the committee's interest, draft proposed recommendations. These are then submitted to the full membership of the Council for careful review and approval before being published.

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THE NCRP

The following comprise the current officers and membership of the Council: Officers President Vice President Secretary and Treasurer Assistant Secretury Assistant Treasurer

CHARLESB. MEINHOLD S.JAMES ADE~~TEIN W. ROGERNEY CARLD. HOBELMAN JAMES F. BERG

Members I. FABRIKANT* WILLIAMA. MILLS SEYMOUR ABRAHAMSON JACOB THOMAS F. GESELL DADEW. MOELLER S. JAMES ADELSTEIN PETER R.ALMOND ETHELS. GILBERT GILBERTS. OMENN LYNNR. ANSPAUGH ROBERTA. GOEPP LESTER J. PETERS JOEL E.GRAY RONALDPETERSEN JOHN A. AUXIER JOHN W. POSTON,SR. ARTHURW. GUY JOHN W. BAUM ERICJ. HALL ANDREWK. POZNANSKI HAROLDL. BECK GENEVIEVE S. ROESSLER NAOMIH. HARLEY MICHAELA. BENDER MARVINROSENSTEIN B. GORDONB L A ~ O C K WILLIAMR. HENDEE BRUCEB. BOECKER DAVIDG. HOEL LAWRENCE N. ROTHENBERG JOHN D. BOICE,JR. F. OWENHOFFMAN MICHAELT. RYAN KEITHJ . SCHIAGER DONALDG. JACOBS AND& BOUVILLE ROBERTL. BRENT A. EVERET~E JAMES, JR. ROBERTA. SCHLENKER JOHN R. JOHNSON ROYE. SHORE A. BERTRAND BRILL ANTONEL. BROOKS BERNDKAHN DAVIDH. SLLNEY KENNETHR. KASE PAULSLOVIC PAULL. CARSON MELVINW. CARTER AMYKRONENBERG RICHARDA. TELL JAMES E. CLEAVER HAROLDL. KUNDEL WILLIAML. TEMPLETON FRED T. CROSS CHARLESE. LAND THOMASS. TENFORDE RALPHH. THOMAS GAILDE PLANQUE JOHN B LITTLE JOHN E. TILL SARAHDONALDSON HARRYR. MAXON CARLH. DURNEY ROGER0. MCCLELLAN ROBERTL. ULLRICH KEITH F. ECKERMAN BARBARAJ . MCNEIL DAVIDWEBER CHARLESM. EISENHAUERCHARLESB. MEINHOLD F. WARDWHICKER THOMASS. ELY FRED A. METIZER,JR. MARVINC. ZISKIN Honorary Members LAURISTONS. TAYLOR,H o ~ o M TPresident ~ WARRENK . SINCLAIR,President Emeritus JOHN H. RUST R.J. MICHAELFRY EDWARD L.ALPEN ROBERT0 . GORSON EUGENEL. SAENGER WILLIAMJ. BAIR JOHN W. HEALY LEONARD A. SAGAN VICTORP. BOND REYNOLD F. BROWN PAULC. HODGES WILLIAM J . SCHULL RANDALLS. CASWELL GEORGEV. LEROY J . NEWELLSTANNARD JOHN B. STORER FREDERICKP. COWAN WILFRIDB. MANN A. ALANMOGHISSI ROY C. THOMPSON JAMES F. CROW GERALDD. DODD KARLZ. MORGAN ARTHURC. UPTON GEORGEL. VOELZ PATRICIAW. DURBIN ROBERTJ. NELSEN MERRILEISENBUD WESLEYL. NYBORG EDWARDW. WEBSI'ER ROBLEYD. EVANS CHESTERR. RICHMOND GEORGEM. WILKENING RICHARD F. FOSTER HARALDH. ROSSI HAROLD 0.WYCKOFF HYMERL. FRIEDELL WILLIAML. RUSSELL 'deceased

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Currently, the following subgroupsare actively engaged in formulating recommendations: Basic Radiation Pmtection Criteria SC 1-3 Collective Dose SC 1-4Extrapolation of Risk from Non-human Experimental Systems to Man SC 1-5 Uncertainty in Risk Estimates Stmctural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV X-Ray Protection in Dental Offices Operational Radiation Safety SC 46-8 Radiation Protection Design Guidelines for Particle Accelerator Facilities SC 46-9 ALARA a t Nuclear Plants SC 46-10 Assessment of Occupational Doses from Internal Emitters SC 46-11 Radiation Protection During Special Medical Procedures SC 46-12Determination of the Effective Dose Equivalent (and Effective Dose) to Workers for External Exposure to Low-LET Radiation Dosimetry and Metabolism of Radionuclides SC 57-2 Respiratory %ct Model SC 57-9 Lung Cancer Risk SC 57-10Liver Cancer Risk SC 57-14 Placental Transfer SC 57-15 Uranium SC 57-16 Uncertainties in the Application of Metabolic Models Radiation Exposure Control in a Nuclear Emergency SC 63-1 Public Knowledge Radionuclides in the Environment SC 64-6 Screening Models SC 64-17 Uncertainty in Environmental Transport in the Absence of Site Specific Data SC 64-18Plutonium Quality Assurance and Accuracy in Radiation Protection Measurements Biological Effects and Exposure Criteria for Ultrasound Efficacy of Radiographic Procedures Radiation Protection in Mammography Guidance on Radiation Received in Space Activities Guidance on Occupational and Public Exposure Resulting from Diagnostic Nuclear Medicine Procedures Radionuclide Contamination SC 84-1 Contaminated Soil SC 84-2 Decontamination and Decommissioning of Facilities Risk of Lung Cancer from Radon Hot Particles in the Eye, Ear or Lung Radioactive and Mixed Waste SC 87-1 Waste Avoidance and Volume Reduction SC 87-2 Waste Classification Based on Risk SC 87-3 Performance Assessment

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SC 88

Fluence as the Basis for a Radiation Protection System for Astronauts SC 89 Nonionizing Electromagnetic Fields SC 89-1 Biological Effects of Magnetic Fields SC 89-2 Practical Guidance on the Evaluation of Human Exposure to Radiofrequency Radiation SC 89-3 Extremely Low-Frequency Electric and Magnetic Fields SC 90 Precautions in the Management of Patients Who have Received Therapeutic Amounts of Radionuclides SC 91 Radiation Protection in Medicine Ad Hoc Committee on the Embryo Fetus and Nursing Child

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 a s follows: American Academy of Dermatology 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 American Nuclear Society American Podiatric Medical Association American Public Health Association

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American Radium Society American Roentgen Ray Society American Society of Radiologic Technologists American Society for Therapeutic Radiology and Oncology Association of University Radiologists Bioelectromagnetics Society College of American Pathologists Conference of Radiation Control Program Directors Electric Power Research Institute Federal Communications Commission Federal Emergency Management Agency Genetics Society of America Health Physics Society Institute of Nuclear Power Operations International Brotherhood of Electrical Workers National Aeronautics and Space Administration National Electrical Manufacturers Association National Institute of Standards and Technology Nuclear Management and Resources Council Oil, Chemical and Atomic Workers Union Radiation Research Society Radiological Society of North America Society of Nuclear Medicine United States Air Force United States Army United States Department of Energy United States Department of Housing and Urban Development United States Department of Labor United States Environmental Protection Agency United States Navy United States Nuclear Regulatory Commission United States Public Health Services Utility Workers Union of America

The NCRP has found its relationships with these organizations to be extremely valuable to continued progress in its program. Another aspect ofthe 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 an opportunity to make suggestions on new studies and related matters. The following organizations participate in the Special Liaison Program:

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Australian Radiation Laboratory Commissariat a 1'Energie Atomique (France) Commission of the European Communities Defense Nuclear Agency Federal Emergency Management Agency International Commission on Non-Ionizing Radiation Protection Japan Radiation Council National Radiological Protection Board (United Kingdom) National Research Council (Canada) Office of Science and Technology Policy Office of Technology Assessment Ultrasonics Institute (Australia) United States Air Force United States Coast Guard United States Department of Health and Human Services United States Department of Transportation United States Nuclear Regulatory Commission

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: Amersham Corporation Commonwealth Edison Consumers Power Company Duke Power Company Eastrnan Kodak Company EG&G Rocky Flats Landauer, lnc. Public Service Electric and Gas Company Southern California Edison Company Westinghouse Electric Corporation 3M

The Council's activities are made possible by the voluntary contribution 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 Oral and Maxillofacial Radiology American Association of Physicists in Medicine American Cancer Society

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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 Pediatric 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 Defense Nuclear Agency Edison Electric Institute Edward Mallinckrodt, Jr. Foundation EG&G Idaho, Inc. Electric Power Research Institute Federal Emergency Management Agency Florida Institute of Phosphate Research 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 National Aeronautics and Space Administration National Association of Photographic Manufacturers National Cancer Institute National Electrical Manufacturers Association National Institute of Standards and Technology Nuclear Management and Resources Council Picker International Radiation Research Society Radiological Society of North America Richard Lounsbery Foundation

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Sandia National Laboratory 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 recommendations based on leading scientificjudgement 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.

NCRP Publications NCRP publications are distributed by the NCRP Publications Office. Information on prices and how to order may be obtained by directing an inquiry to: NCRP Publications 7910 Woodmont Avenue Suite 800 Bethesda, MD 20814-3095 The currently available publications are listed below.

No. 8

NCRP Reports Title Control and Removal ofRadioactive Contamination in Laboratories (1951) Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure (1959)[Includes Addendum 1 issued in August 19631 Measurement of Neutron Flux and Spectra for Physical and Biological Applications (1960) Measurement of Absorbed Dose ofNeutrons,and of Mixtures of Neutrons and Gamma Rays (1961) Stopping Powers for Use with Cavity Chambers (1961) Safe Hand1ing of Radioactive Materials (1964) Radiation Protection in Educational Institutions (1966) Dental X-Ray Protection (1970) Radiation Protection in Veterinary Medicine (1970) Precautions in the Management of Patients Who Have Received Therapeutic Amounts of Radionuclides (1970) Protection Against Neutron Radiation (1971) Protection Against Radiation from Brachytherapy Sources (1972) Specification of Gamma-Ray Brachytherapy Sources (1974) Radiological Factors Affecting Decision-Making i n a Nuclear Attack (1974)

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NCRP PUBLICATIONS

Krypton-85 in the Atmosphere-Accumulation, Biological Significance, and Control Technology (1975) Alpha-Emitting Particles in Lungs (1975) Tritium Measurement Techniques (1976) Structural Shielding Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies Up to 10 MeV (1976) Environmental Radiation Measurements (1976) Radiation Protection Design Guidelines for 0.1-100 MeV Particle Accelerator Facilities (1977) Cesium-137 from the Environment to Man: Metabolism and Dose (1977) Medical Radiation Exposure of Pregnant and Potentially Pregnant Women (1977) Protection of the Thyroid Gland in the Event of Releases of Radioiodine (1977) Instrumentation and Monitoring Methods for Radiation Protection (1978) A Handbook of Radioactivity Measurements Procedures, 2nd ed. (1985) Operational Radiation Safety Program (1978) Physical, Chemical, and Biological Properties of Radiocerium Relevant to Radiation Protection Guidelines (1978) Radiation Safety Training Criteria for InduWid Radiography (1978) Tritium in the Environment (1979) Tritium and Other Radionuclide Labeled Organic Compounds Incorporated in Genetic Material (1979) Influence of Dose and Its Distribution in Time on DoseResponse Relationships for Low-LET Radiations (1980) Management of Persons Accidentally Contaminated with Radionuclides (1980) Radiofrequency Electromagnetic Fields-Properties, Quantities and Units, Biophysical Interaction, and Measurements (1981) Radiation Protection in Pediatric Radiology (1981) Dosimetry of X-Ray and Gamma-Ray Beams for Radiation Therapy in the Energy Range 10 keV to 50 MeV (1981) Nuclear Medicine-Factors Influencing the Choice and Use of Radionuclides i n Diagnosis and Thempy (1982) Opemtional Radiation Safety-Training (1983) Radiation Protection and Measurement for Low -Voltage Neutron Generators (1983)

NCRP PUBLICATIONS

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Protection in Nuclear Medicine and Ultrasound Diagnostic Procedures in Children (1983) Biological Effects of Ultrasound: Mechanisms and Clinical Implications (1983) Iodine-129: Evaluation of Releases from Nuclear Power Genemtion (1983) Radiological Assessment: Predicting the Transport, Bioaccumulation, and Uptake by Man of Radionuclides Released to the Environment (1984) Exposures from the Uranium Series with Emphasis on Radon and Its Daughters (1984) Evaluation of Occupational and Environmental Exposures toRadon and Radon Daughters in the United States (1984) Neutron Contamination from Medical Electron Accelemtors (1984) Induction of Thyroid Cancer by Ionizing Radiation (1985) Carbon-14 in the Environment (1985) SI Units in Radiation Protection and Measurements (1985) The Experimental Basis for Absorbed-Dose Calculations in Medical Uses of Radionuclides (1985) General Concepts for the Dosimetry of Internally Deposited Radionuclides (1985) Mammography-A User's Guide (1986) Biological Effects and Exposure Criteria for Radiofrequency Electromagnetic Fields (1986) Use of Bioassay Procedures for Assessment c. Internal Radionuclide Deposition (1987) Radiation Alarms and Access Control Systems (1986) Genetic Effects from Internally Deposited Radionuclides (1987) Neptunium: Radiation Protection Guidelines (1988) Public Radiation Exposure from Nuclear Power Generation in the United States (1987) Ionizing Radiation Exposure of the Population of the United States (1987) Exposure of the Population in the United States and Canada from Natural Background Radiation (1987) Radiation Exposure of the U.S. Population from Consumer Products and Miscellaneous Sources (1987) Comparative Carcinogenicity of Ionizing Radiation and Chemicals (1989) Measurement of Radon and Radon Daughters in Air (1988) Guidance on Radiation Received in Space Activities (1989) Quality Assurance for Diagnostic Imaging (1988)

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Exposure of the U.S. Population from Diagnostic Medical Radiation (1989) 101 Exposure of the U.S. Population from Occupational Radiation (1989) 102 Medical X-Ray, Electron Beam and Gamma-Ray Protection for Energies Up to 50 MeV (Equipment Design, Performance and Use) (1989) 103 Control of Radon in Houses (1989) 104 The Relative Biological Effectiveness ofRadiations ofDifferent Quality (1990) 105 Radiation Protection for Medical a n d Allied Health Personnel (1989) 106 Limit for Exposure to "Hot Particles" on the Skin (1989) 107 Implementation of the Principle of As Low As Reasonably Achievable (ALARA) for Medical and Dental Personnel (1990) Conceptual Basis for Calculations of Absorbed-Dose 108 Distributions (1991) 109 Effects of Ionizing Radiation on Aquatic Organisms (1991) 110 Some Aspects of Strontium Radiobiology (1991) 111 Developing Radiation Emergency Plans forAcademic, Medical or Industrial Facilities (1991) 112 Calibration of Survey Instruments Used in Radiation Protection for the Assessment of Ionizing Radiation Fields and Radioactive Surface Contamination (1991) 113 Exposure Criteria for Medical Diagnostic Ultrasound:I. Criteria Based on Thermal Mechanisms (1992) 114 Maintaining Radiation Protection Records (1992) 115 Risk Estimates for Radiation Protection (1993) 116 Limitation of Exposure to Ionizing Radiation (1993) 117 Research Needs for Radiation Protection (1993) 118 Radiation Protection i n the Mineral Extraction Industry (1993) 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. 32-118).Each binder will accommodate from five to seven reports. The binders carry the identification "NCRP Reports" and come with label holders which permit the user to attach labels showing the reports contained in each binder. The following bound sets of NCRP reports are also available: Volume I. NCRP Reports Nos. 8,22 Volume 11. NCRP Reports Nos. 23,25,27,30 Volume 111. NCRP Reports Nos. 32,35,36,37 100

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Volume IV. NCRP Reports Nos. 38,40,41 Volume V. NCRP Reports Nos. 42,44,46 Volume VI. NCRP Reports Nos. 47,49,50,51 Volume VII. NCRP Reports Nos. 52,53,54,55,57 Volume VIII. NCRP Report No. 58 Volume IX. NCRP Reports Nos. 59,60,61,62,63 Volume X. NCRP Reports Nos. 64,65,66,67 Volume XI. NCRP Reports Nos. 68,69,70,71,72 Volume XII. NCRP Reports Nos. 73,74,75,76 Volume XIII. NCRP Reports Nos. 77,78,79,80 Volume XIV. NCRP Reports Nos. 81,82,83,84,85 Volume XV. NCRP Reports Nos. 86,87,88,89 Volume XVI. NCRP Reports Nos. 90,91,92,93 Volume XVII. NCRP Reports Nos. 94,95,96,97 Volume XVIII. NCRP Reports Nos. 98,99,100 Volume XIX. NCRP Reports Nos. 101,102,103,104 Volume XX. NCRP Reports Nos. 105,106,107,108 Volume XXI. NCRP Reports Nos. 109,110,111 Volume XXII. NCRP Reports Nos. 112,113,114 (Titles of the individual reports contained in each volume are given above.)

NCRP Commentaries No. 1

Title Kryljton85 in the Atmosphere--With Specific Reference to the Public Health Significance of the Proposed Controlled Release at Three Mile Island (1980) Preliminary Evaluation of Criteria for the Disposal of Transuranic Contaminated Waste (1982) Screening Techniques for Determining Compliance with Environmental Standards-Releases of Radionuclides to the Atmosphere (19861,Revised (1989) Guidelines for the Release of Waste Water from Nuclear Facilities with Special Reference to the Public Health Significance of the Proposed Release of Treated Waste Waters at Three Mile Island (1987) Review of the Publication, Living Without Landfills (1989) Radon Exposure of the U.S. Population-Status of the Problem (1991) Misadministration of Radioactive Material in MedicineScientific Background (1991)

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Uncertainty in NCRP Screening Models Relating to Atmospheric Transport, Deposition and Uptake by Humans (1993)

Proceedings of the Annual Meeting No. 1

Title Perceptions of Risk, Proceedings of the Fifteenth Annual Meeting held on March 14-15, 1979 (including Taylor Lecture No. 3) (1980) Critical Issues in Setting Radiation Dose Limits, Proceedings of the Seventeenth Annual Meeting held on April 89, 1981 (including Taylor Lecture No. 5) (1982) Radiation Protection and New Medical Diagnostic Approaches, Proceedings of the Eighteenth Annual Meeting held on April 6-7, 1982 (including Taylor Lecture No. 6) (1983) Environmental Radioactivity, Proceedings of the Nineteenth Annual Meeting held on April 6-7,1983 (including Taylor Lecture No. 7) (1983) Some Issues Important in Developing Basic Radiation Protection Recommendations, Proceedings of the Twentieth Annual Meeting held on April 4-5,1984 (includingTaylor Lecture No. 8) (1985) Radioactive Waste, Proceedings of the Twenty-first Annual Meeting held on April 3-4,1985 (includingTaylor Lecture No. 9) (1986) Nonionizing Electromagnetic Radiations and Ultmsound, Proceedings of the Twenty-second Annual Meeting held on April 2-3, 1986 (including Taylor Lecture No. 10) (1988) New Dosimetry a t Hiroshima and Nagasaki and Its Implications for Risk Estimates, Proceedings of the Twenty-third Annual Meeting held on April 8-9,1987 (includingTaylor Lecture No. 11)(1988) Radon, Proceedings of the Twenty-fourth Annual Meeting held on March 30-31, 1988 (including Taylor Lecture No. 12) (1989) RadiationProtection Today--The NCRP a t Sixty Years, Proceedings of the Twenty-Mth Annual Meeting held on April 5-6, 1989 (including Taylor Lecture No. 13) (1990) Health and Ecological Implications of Radioactively Contaminated Environments, Proceedings of the Twenty-

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sixth Annual Meeting held on April 4-5,1990 (including Taylor Lecture No. 14) (1991) Genes, Cancer and Radiation Protection, Proceedings of the Twenty-seventh Annual Meeting held on April 3-4,1991 (including Taylor Lecture No. 15) (1992) Radiation Protection in Medicine, Proaxdings of the Twentyeighth Annual Meeting held on April 1-2,1992 (including Taylor Lecture No. 16) (1993) Lauriston S. Taylor Lectures

No. 1 2

Title The Squares of the Natural Numbers inRadiation Protection by Herbert M. Parker (1977) Why be Quantitative about Radiation Risk Estimates? by Sir Edward Pochin (1978) Radiation Protection--Concepts and Trade Offs by Hymer L. Friedell (1979) [Available also in Perceptions of Risk, see abovel From "Quantity of Radiation" and "Dose" to "Exposure" and "Absorbed Dose9'-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 CriticalIssues in Setting Radiation Dose Limits, see abovel Ethics, Trade-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 in Radiation Protection by Harald H . Rossi (1984) [Available also in Some Issues Important in Developing Basic Radiation Protection Recommendations, see above] Truth (and Beauty) in Radiation Measurement by John H. Harley (1985) [Available also in Radioactive Waste, see above] Biological Effects of Non-ionizing Radiations: Cellular Properties and Interactions by Herman P. Schwan (1987) [Available also in Nonionizing Electromagnetic Radiations and Ultrasound, see abovel How to be Quantitative about Radiation Risk Estimates by

12 13 14

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Seymour Jablon (1988) [Available also in New Dosimetry at Hiroshima and Nagasaki and its Implications for Risk Estimates, see abovel How Safe is Safe Enough? by Bo Lindell(1988) [Available also in Radon, see abovel Radiobiology and Radiation 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 above] Dose and Risk i n Diagnostic Radiology: How Big? How Little? by Edward W. Webster (1992)[Available also in Radiation Protection in Medicine, see abovel

Symposium Proceedings The Control of Exposure of the Public to Ionizing Radiation i n the Event ofAccident or Attack, Proceedings of a Symposium held April 27-29, 1981 (1982) NCRP Statements No. 1 2

3 4 5 6

Title "Blood Counts, Statement of the National Committee on Radiation Protection," Radiology 63, 428 (1954) ''Statements on Maximum Permissible Dose from Television Receivers and Maximum Permissible Dose to the Skin of the Whole Body," Am. J. Roentgenol., Radium Ther. and Nucl. Med. 84, 152 (1960) and Radiology 75, 122 (1960) X-Ray Protection Standards for Home Television Receivers, Interim Statement of the National Council on Radiation Protection and Measurements (1968) Specification of Units ofNatura1 Uranium and Natural Thorium, Statement of the National Council on Radiation Protection and Measurements, (1973) NCRP Statement on Dose Limit for Neutrons (1980) Control of Air Emissions of Radionuclides (1984)

NCRP PUBLICATIONS

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The Probability That a Particular Malignancy May Have Been Caused by a Specified Zrmdiation (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) The following documents are now superseded andlor out of print:

NCRP Reports No.

Title X-Ray Protection (1931) [Superseded by NCRP Report No. 31 Radium Protection (1934) [Superseded by NCRP Report No. 41 X-Ray Protection (1936) [Superseded by NCRP Report No. 61 Radium Protedion (1938) [Supersededby NCRP Report No. 131 Safe Handling of Radioactive Luminous Compound (1941) [Out of Printl Medical X-Ray Protection Up to Two Million Volts (1949) [Superseded by NCRP Report No. 181 Safe Handling of Radioactive Isotopes (1949) [Superseded by NCRP Report No. 301 Recommendations for WasteDisposal ofPhosphorus32 and Zodine-131 for Medical Users (1951) [Out of Printl Radiological Monitoring Methods and Instruments (1952) [Superseded by NCRP Report No. 571 Maximum Permissible Amounts of Radioisotopes i n the Human Body and Maximum Permissible Concentrations in Air and Water (1953) [Superseded by NCRP Report No. 221 Recommendations for the Disposal of Carbon-14 Wastes (1953) [Superseded by NCRP Report No. 811

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Protection Against Radiations from Radium, Cobalt-60 and Cesium-137 (1954)[Superseded by NCRP Report No. 241 Protection Against Betatron-Synchrotron Radiations Up to 100 Million Electron Volts (1954)[Superseded by NCRP Report No. 511 Safe Handling of Cadavers Containing Radioactive Isotopes (1953)[Superseded by NCRP Report No. 211 Radioactive-Waste Disposal i n the Ocean (1954)[Out of Print] Permissible Dose from External Sources of Ionizing Radiation (1954)including Maximum Permissible Exposures to Man, Addendum to National Bureau of Standards Handbook 59 (1958)[Superseded by NCRP Report No. 391 X-Ray Protection (1955)[Superseded by NCRP Report No. 261 Regulation of Radiation Exposure by Legislative Means (1955)[Out of Print] Protection Against Neutron Radiation Up to 30 Million Electron Volts (1957)[Superseded by NCRP Report No. 381 Safe Handling of Bodies Containing Radioactive Isotopes (1958)[Superseded by NCRP Report No. 371 Protection Against Radiations from Sealed Gamma Sources (1960)[Superseded by NCRP Reports No. 33,34and 401 Medical X-Ray Protection Up to Three Million Volts (1961) [Superseded by NCRP Reports No. 33,34,35and 361 A Manual of Radioactivity Procedures (1961)[Superseded by NCRP Report No. 581 Exposure to Radiation in an Emergency (1962)[Superseded by NCRP Report No. 421 Shielding for High-Energy Electron Accelerator Installations (1964)[Superseded by NCRP Report No. 511 Medical X-Ray and Gamma-Ray Protection for Energies up to 10 MeV-Equipment Design and Use (1968)[Superseded by NCRP Report No. 1021 Medical X-Ray and Gamma-Ray Protection for Energies Up to 10 MeV-Structural Shielding Design and Evaluation Handbook (1970)[Superseded by NCRP Report No. 491 Basic Radiation Protection Criteria (1971)[Superseded by NCRP Report No. 911 Review of the Current State of Radiation Protection Philosophy (19751[Superseded by NCRP Report No. 911 Natural Background Radiation in the United States (1975) [Superseded by NCRP Report No. 941

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Radiation Protection for Medical and Allied Health Personnel (1976) [Superseded by NCRP Report No. 1051 Review ofNCRP RadiationDose Limit forEmbryo and Fetus in Occupationally-Exposed Women (1977) [Out of Print] Radiation Exposure from Consumer Products and Miscellaneous Sources (1977) [Superseded by NCRP Report No. 951 A Handbook of Radioactivity Measurements Procedures, 1st ed. (1978) [Superseded by NCRP Report No. 58,2nd ed.] Mammography (1980) [Out of Print] Recommendations on Limits for Exposure toIonizing Radiation (1987) [Superseded by NCRP Report No. 1161

NCRP Proceedings No. 2

Title Quantitative Risk in Standards Setting, F'roceedings of the Sixteenth Annual Meeting held on April 2-3, 1980 [Out of Print]

Index Access control, 27, 31-32, 80, 83, 85 Airborne radioactivity, 19-21 Airborne radionuclides (longlived), 9, 15, 36, 42-44, 62, 64, 80, 82, 84, 88, 89, 91, 94 Airborne radon and radon progeny, 25-26,44-46,63-64 Alpha-specific survey probe, 47 Aluminum, 16 Ammonium diuranate, 82 Ammonium uranyl tricarbonate, 81

Area monitoring, 39,43,45,47 As low as reasonably achievable (ALARA), 3,22-24,27,30,32, 48, 70, 72, 89, 90, 101 Beneficiation, 86, 101 Beta radiation fields, 19, 24, 84 Bioassay, 10, 49-52, 86, 101, 102 Breathing zone sampling, 44 By-product material, 68, 70 By-products, 17, 22, 71, 80, 90 Calcining, 21, 82, 83, 94 Changelshower rooms, 34,47,83 Chemical element influences, 53, 65 Consultants, outside assistance, 6, 40, 56, 76 Contamination control, 20, 27, 30, 33, 47, 48 Contamination, surface, 9, 20, 22, 47,92, 101 Copper, 16,81 Crushing and grinding, 19-21, 25, 29, 32, 42 Decommissioning, 4, 7 Design of facility, 10, 26, 90 Dilution, dispersion, 54.56, 60 Dissolved fraction in water, 58, 64

Dose determinants, general, 18, 24 Dosimetry accreditation program, 41 Dry rock handling, 86, 89 Effluent control systems, 29, 30, 53,55, 59, 83 Effluent monitoring, 53,55-58, 61,62, 89 air monitoring, 56-58 frequency, 57, 58 objectives, 58-59 program design, 59-63 water monitoring, 58-59 Effluent monitoring and environmental surveillance, 53-65 effluent monitoring, 55-58 environmental pathways, 54-55 environmental surveillance, 58-65 Effluents, 10, 20, 21, 53 Elevated radiation levels, 17, 88, 92 Emergency preparedness and response, 62-63, 75, 83 Enhanced radiation levels, 17, 25 Environmental impact assessment, 10,55, 56,81, 93 modeling, 10,56, 59 Environmental monitoring, 53, 58-61, 78,80,84 baseline, 53,59-61 desirability, 60 heap leaching, 78 in situ facility, 80 objectives, 58-59 side-stream extraction facility, 84 Environmental surveillance, 58-65

INDEX

Equilibrium, 15, 25, 44, 85, 86, 90,102 Equipment and system design, 28-31,80,85,87 Exposure environment, 24 Exposure limits, personnel, 23, 38,48, 59, 70, 72 Exposure management program, 23-35 employee training, 34, 35 exposure environment, 2 4 2 6 exposure limits, 23, 24 facility design and engineering, 26-31 facility procedures and practices, 31-34 Exposure pathways, 14,54 External radiation sources, 18, 19, 22, 24, 29, 38-42, 65 Extractive processes, 15, 17 Facility layout, 27-28 Ferrophosphorus, 94 Film badges, 40, 41 Filter pan repair, 91 Fluorimetry, 43 Fluorospar, 16 Gamma radiation field, 19, 24, 39, 61, 79,80, 82, 84, 88, 90, 91, 93, 94 environmental measurements, 61,65 Gauges, 19,32,88 Geiger-Mueller meters, 39 Green salt (UF,), 82 Guidance and consensus standards, 66, 67 Guidelines, standards and regulations, 66-74 approaches to limits, 70-71 effluents and the environment, 73-74 occupational exposure, 71-73 public exposure, 73-74 rules and regulations, 68-70 sources, 67-68

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137

Heap leaching, 77-79, 102 Housekeeping, 32, 80

In situ (solution) mining, 77, 79-80 Instrument calibration, 40 Ion chamber meters, 40 Iron, 16 Isokinetic sampling, 57, 58 Job analysis, 34-35 Layout of facility, 27,80,82, 85 Lead, 15 Linear no-threshold hypothesis, 2 Lower limits of detection, 57, 64 Lunch rooms, 20,27,47 Maintenance shops, 27,47,91 Management principles, 7, 10, 23, 55, 58-59, 61 Meteorological data, 56 Mill exhaust stacks, 21, 55, 57 Mine ventilation, 21, 28, 46, 56-57 Mine waters, 30,55, 79 Molybdenum, 16,81 Monazite, 16,84 Monitoring program, 37-38 Natural gas, 15, 16, 77 Naturally occurring radioactive materials, 14-18 Natural radiation environment, 1, 14, 103 Niobium, 16 Occupational exposure monitoring, 36-52 airborne radon and progeny, 44-46 bioassay, 49-52 external radiation, 38-42 long-lived airborne radionuclides, 42-44 monitoring program, 37-38 surface contamination, 47-48

138

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INDEX

Open pit or surface mines, 27, 29, 57, 86, 88 Ore dust, 25, 29, 42, 47, 58, 78 Ore stockpiles, 21, 27, 29 Particle sizes, 18, 64 Personal monitoring, 37, 40, 43-45, 48, 49, 79, 80, 82, 83 Personal protective equipment, 33,80, 83, 85, 89 Personal samplers, 44 Personnel monitoring (see personal monitoring) Personnel protective equipment, 33-34 Petroleum, 15, 16, 77 Phosphate, 1 6 , 8 6 9 5 mining, beneficiation and wet rock handling, 86-87 occupational exposure, 87-88, 89, 90, 91 phosphate rock drying and dry handling, 89-90 production of phosphate products, 93-94 wet process phosphoric acid plants, 90-94 Phosphate rock, 87, 103 Phosphogypsum, 90,93-94 Phosphoric acid, 81, 86 Phosphorus (thermal process), 86 Potash, 16 Procedures and practices, 10, 31-34, 48, 66,80, 83, 85 Product packaging, 20,21, 25,29, 42-44, 82 Program documentation, 10 Protective clothing, 33-34 Purpose of report, 1 Quality assurance program, 10-12,41, 56,58 Radiation conditions documentation, 9, 10 Radiation emergency response planning, 75-76 environment, 76

operations, 75 transportation, 76 Radiation exposure management, 23,26 Radiation monitoring, 36, 37, 40, 43, 45, 46, 48, 49 personnel exposure assessment, 36, 37, 40, 43, 45,48, 49 workplace characterization, 36, 43, 45, 46 Radiation protection organization, 5, 6 Radiation protection program, design of, 5-13 criteria for, 5-6 integration into overall safety program, 12-13 management of, 6-7 quality assurance, 11-12 records, preparation and maintenance of, 9-11 Radiation safety committee, 8 Radiation safety officer, 8, 92 Radiation safety role, 7 Radioactive decay series, 14, 85, 86,96-99,103 Radioactive material control, 32 Radium, 16-18, 25,28, 53, 64, 80,81,88,90,92-94 Radon, 9, 15, 19, 25, 28, 30, 44-46, 63, 80, 84,87, 91, 93, 94,104 Radon flux, 63, 93,104 Radon monitor, etched track, 63 Radon progeny, 9, 15, 19,25, 28, 29,44, 46, 63, 85,87, 94, 104 Rare earths, 16,84 Recordkeeping, 9 Regulations, 5, 66, 68, 72, 73, 88 Release for unrestricted use, 22, 31,48, 92, 93 Residues in equipment, 25, 39, 72,80,90-92 Respirators, 33, 48,49, 89 Respiratory protection, 33 Reuse and salvage, 17, 22, 84 Risk, 2, 5, 12 Rutile, 84

INDEX

Safety program, overall, 12 Sample preservation, 65 Sampling, 36, 43, 45,55, 58, 61-65, 78, 80, 81, 87, 89, 93, 104 Sanitary or storm sewers, 56 Sealed source control, 32 Scintillation meters, 39 Screening models, 59-60 Settling pond, 21, 31 Side-stream extraction, 77, 80-84,104 Silica, 12 Site selection, 26-27 Slag, 94 Soil and vegetation, monitoring and surveillance, 64 Solubility, 18 Source material, 68, 70 Sources of exposure, 14-22 characterization, 14-18 environmental releases, 20-22 occupational, 18-20 process by-products, 22 waste, 22 Surface contamination, 20, 47-48 Surveys, 9, 36, 37, 39, 47, 85, 88, 90,92 Suspended fraction in water, 58, 64

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139

Tailings, 30, 80, 84, 87, 104 Thermal process (phosphorus), 94 Thermoluminescent dosimeters, 40, 41, 63, 65, 104 Thorium, 14, 16, 18, 68, 74, 84, 87,104 Thorium and rare earths processing, 84-86 Tin, 16,81 Titanium, 16 Training, 10, 32,34, 75, 80, 85 Tungsten, 81 Uranium, 4, 12, 14, 16, 18, 24, 43, 49, 68, 74, 80, 81, 84, 87, 96 Vanadium, 16 Waste management, 32 Wastes, 18, 22, 30, 32, 84, 87, 92 Waterborne effluents, 21,64-65 Wet process phosphoric acid, 90 Wet rock handling, 86, 87 Workers role in safety program, 7 Working levels, 26, 104 Workplace characterization, 36, 43,45, 46 Zinc, 15, 16 Zirconium, 16