DOE-STD-1136-2004, Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities

NOT MEASUREMENT SENSITIVE DOE-STD-1136-2004 December 2004 DOE STANDARD GUIDE OF GOOD PRACTICES FOR OCCUPATIONAL RADIOLOGICAL PROTECTION IN URANIUM FACILITIES U.S. Department of Energy AREA SAFT Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U. S. Department of Energy, (800) 473-4373, fax (301) 903-9823. Available to the public from the U. S. Department of Commerce, Technology Administration, National Technical Information Service, Springfield, VA 22161; (703) 605-600 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Foreword This Technical Standard (TS) discusses, but does not establish any, requirements for DOE uranium facilities. Its purpose is to provide information that will assist DOE and DOE-contractor health and safety professionals in developing programs that will provide an appropriate level of protection to both affected workers and members of the public affected by DOE uranium-handling activities. This TS provides guides to good practice, update existing reference material, and discuss practical lessons learned relevant to the safe handling, processing, and storage of uranium. The technical rationale for the guidance provided herein is explained to allow affected individuals to adapt the recommendations to similar situations throughout the DOE complex. This TS provides information to assist uranium facilities in complying with Title 10 of the Code of Federal Regulations, Part 835 (10 CFR 835), Occupational Radiation Protection and various DOE Orders, and supplements DOE’s 441.1 series of Guides and DOE-STD-1098-99, Radiological Control. This TS does not include every requirement applicable to DOE uranium facilities. Individuals responsible for developing and implementing radiation protection programs at uranium facilities should be knowledgeable of the requirements that apply to their facilities. Copies of electronic files of this TS may be obtained from the DOE Office of Worker Protection Policy and Programs Home Page Internet site (http://www.eh.doe.gov/radiation/ts.html). Copies of this TS are also available from the DOE Technical Standards Program Internet site (http://www.eh.doe.gov/techstds/). iii DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities This page intentionally left blank. iv DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities TABLE OF CONTENTS CHAPTER 1 - INTRODUCTION 1.0 INTRODUCTION .....................................................................................................................1-1 1.1 PURPOSE AND APPLICABILITY .........................................................................1-1 1.2 DEFINITIONS ........................................................................................................1-1 1.3 DISCUSSION .........................................................................................................1-1 CHAPTER 2 - PROPERTIES AND RELATIVE HAZARDS 2.0 PROPERTIES AND RELATIVE HAZARDS ..............................................................................2-1

2.1 NUCLEAR PROPERTIES OF URANIUM ......................................................................2-1 2.1.1 Isotopic Characterization ..................................................................................2-1 2.1.2 Decay Chains ..................................................................................................2-4 2.1.3 Enrichment ......................................................................................................2-7 2.1.4 Contaminants from Recycled Uranium and Associa ted Hazards ..........................2-11 2.2 PHYSICAL AND CHEMICAL PROPERTIES ................................................................2-16 2.2.1 Uranium Fuel Processing .................................................................................2-17 2.2.2 Uranium Metal ................................................................................................2-17 2.3 RADIOLOGICAL CHARACTERISTICS AND EFFECTS ..............................................2-18 2.3.1 Alpha-Neutron External Hazard ........................................................................2-19 2.3.2 Mode of Uranium Entry into the Body ..............................................................2-20 2.4 CHEMICAL TOXICITY ................................................................................................2-21 2.4.1 Human Response Indicators .............................................................................2-24 2.4.2 Transfer to the Fetus ........................................................................................2-25 2.5 CHEMICAL VERSUS RADIOLOGICAL HAZARDS ....................................................2-25 2.6 INDUSTRIAL HAZARDS .............................................................................................2-31 2.6.1 Hydrogen Fluoride ...........................................................................................2-31 2.6.2 Nitric Compounds ...........................................................................................2-31 2.6.3 Hydrogen Gas .................................................................................................2-31 2.6.4 Fire .................................................................................................................2-32 CHAPTER 3 - RADIATION PROTECTION 3.0 RADIATION PROTECTION .....................................................................................................3-1 3.1 REGULATION AND STANDARDS ..............................................................................3-1 3.2 RADIATION PROTECTION PROGRAMS ....................................................................3-1 3.2.1 Organization and Administration ......................................................................3-1 3.2.2 ALARA Program .............................................................................................3-6 3.2.3 External Dosimetry Program ............................................................................3-9 3.2.4 Internal Dosimetry Program .............................................................................3-9 3.2.5 Area Monitoring and Control ............................................................................3-9 3.2.6 Radiological Controls ......................................................................................3-16 3.2.7 Emergency Exposure Situations ........................................................................3-18 3.2.8 Nuclear Accident Dosimetry .............................................................................3-18 3.2.9 Records ...........................................................................................................3-18 3.2.10 Radiation Safety Training ...............................................................................3-19

v DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities TABLE OF CONTENTS (continued) 3.3 RELATED PROGRAMS ................................................................................................3-21 3.3.1 Onsite Packaging and Transportation ................................................................3-21 3.3.2 Conduct of Operations .....................................................................................3-23 3.3.3 Integrated Safety Management .........................................................................3-25 CHAPTER 4 - CONTAMINATION CONTROL 4.0 CONTAMINATION CONTROL ................................................................................................4-1 4.1 AIR MONITORING ...............................................................................................4-1 4.1.1 Internal Versus External Dose Philosophy .........................................................4-1 4.1.2 Purpose of Air Monitoring ...............................................................................4-2 4.1.3 Regulations and Limits ....................................................................................4-3 4.1.4 Theoretical Considerations and Uncertainties ....................................................4-4 4.1.5 Samplers and Instrumentation ...........................................................................4-8 4.1.6 Sample Activity Measurement ..........................................................................4-11 4.1.7 Continuous Air Monitors ..................................................................................4-12 4.1.8 Monitoring Strategies and Protocols ..................................................................4-12 4.2 SURFACE CONTAMINATION CONTROL ...................................................................4-14 4.2.1 Reporting and Documenting Contamination Levels ............................................4-15 4.2.2 Monitoring ......................................................................................................4-16 4.2.3 Release Criteria ...............................................................................................4-21 4.2.4 ALARA Guidelines .........................................................................................4-24 4.3 PERSONNEL CONTAMINATION CONTROL ..............................................................4-25 4.3.1 Monitoring Philosophy .....................................................................................4-25 4.3.2 Monitoring Program ........................................................................................4-25 4.3.3 Protective Clothing ..........................................................................................4-26 4.3.4 Respiratory Protection ......................................................................................4-26 4.3.5 ALARA Guidelines .........................................................................................4-27 4.3.6 Release Criteria ...............................................................................................4-27 4.4 DECONTAMINATION AND DECOMMISSIONING TECHNIQUES .............................4-27 4.4.1 Personnel Decontamination ..............................................................................4-27 4.4.2 Equipment and Surface Decontamination ..........................................................4-28

CHAPTER 5 - INTERNAL DOSIMETRY 5.0 INTERNAL DOSIMETRY .........................................................................................................5-1 5.1 INTERNAL DOSE EVALUATION PROGRAM .............................................................5-1 5.1.1 Performance Capabilities for Internal Exposure Monitoring ................................5-2 5.1.2 Protection of the Embryo/Fetus, Minors, and Members of the Public ...................5-7 5.2 CHARACTERIZATION OF INTERNAL HAZARDS .....................................................5-8 5.3 SCOPE OF BIOASSAY PROGRAM ..............................................................................5-9 5.3.1 Classification of Bioassay Measurements ..........................................................5-10 5.3.2 Monitoring Requirements and Selection of Employees .......................................5-11 5.3.3 Selection of Bioassay Monitoring Techniques ...................................................5-12 5.4 ESTABLISHING BIOASSAY FREQUENCY .................................................................5-14 5.4.1 Frequency Based on Program Sensitivity ...........................................................5-14 5.4.2 Frequency Based on Potential Risk of Intake .....................................................5-16 v i DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities TABLE OF CONTENTS (continued) 5.4.3 Special Bioassay as Supplements to Routine Bioassay Programs ........................5-16 5.4.4 Long-term Follow-up Bioassay Programs ..........................................................5-17 5.4.5 Other Frequency Situations ..............................................................................5-17 5.5 ADMINISTRATION OF A BIOASSAY PROGRAM ......................................................5-17 5.5.1 In Vivo Monitoring ..........................................................................................5-18 5.5.2 Urine Sampling ...............................................................................................5-19 5.5.3 Fecal Sampling ................................................................................................5-21 5.5.4 Conditions for Adjustments of Action Levels ....................................................5-22 5.6 MODELING THE BEHAVIOR OF URANIUM IN THE BODY ......................................5-22 5.6.1 Respiratory Tract .............................................................................................5-22 5.6.2 Gastrointestinal Tract .......................................................................................5-22 5.6.3 Systemic Retention and Excretion of Uranium ...................................................5-23 5.6.4 Chemical Toxicity ...........................................................................................5-28 5.6.5 Natural Uranium Balance in Man ......................................................................5-29 5.6.6 Mother-to-Fetus Transfer .................................................................................5-30 5.7 INTERPRETATION OF BIOASSAY RESULTS .............................................................5-30 5.7.1 In Vivo Count Results ......................................................................................5-31 5.7.2 Urine Sample Results .......................................................................................5-32 5.7.3 Fecal Sample Results .......................................................................................5-33 5.7.4 Use of Air Sample Data in Internal Dosimetry ...................................................5-33

5.8 DOSE ASSESSMENT ....................................................................................................5-34 5.8.1 Methods of Estimating Intake ...........................................................................5-35 5.8.2 Alternate Methods of Intake Assessment ...........................................................5-36 5.8.3 Estimating Effective Dose Equivalent from Intakes of Uranium ..........................5-36 5.9 REFERENCE AND ACTION LEVELS ..........................................................................5-37 5.10 RESPONSE TO SUSPECTED INTAKES .....................................................................5-39 5.10.1 Emergency Action Planning ...........................................................................5-41 5.10.2 Medical Emergency Response Plan .................................................................5-41 5.10.3 Responsibilities for Management of Internal Contamination .............................5-41 5.10.4 Immediate Medical Care ................................................................................5-42 5.10.5 Contaminated Wounds ...................................................................................5-42 CHAPTER 6 - EXTERNAL DOSIMETRY 6.0 EXTERNAL DOSIMETRY ........................................................................................................6-1 6.1 DOSE LIMITS ..............................................................................................................6-1 6.1.1 Limiting Quantities ..........................................................................................6-1 6.1.2 Operational Quantities .....................................................................................6-3 6.2 RADIATIONS IN URANIUM FACILITIES ...................................................................6-4 6.2.1 Alpha and Beta Doses ......................................................................................6-7 6.2.2 Gamma Doses .................................................................................................6-8 6.2.3 Neutron Dose Equivalents ................................................................................6-8 6.3 RADIATION DETECTION AND EVALUATION ..........................................................6-10 6.3.1 Portable Survey Instruments--Beta Radiation Response ......................................6-11 6.3.2 Portable Survey Instruments--Gamma Radiation Response .................................6-14 6.3.3 Portable Survey Instruments--Neutron Response ...............................................6-17 vii DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities TABLE OF CONTENTS (continued) 6.4 PERSONNEL DOSIMETRY ..........................................................................................6-18 6.4.1 Energy Dependence .........................................................................................6-18 6.4.2 Angular Dependence ........................................................................................6-19 6.4.3 Dosimetry Practices .........................................................................................6-19 6.4.4 Extremity Dosimetry ........................................................................................6-20 6.4.5 Dose to Lens of Eye .........................................................................................6-22

6.5 EXTERNAL DOSE CONTROL .....................................................................................6-22 6.5.1 Time ...............................................................................................................6-22 6.5.2 Distance ..........................................................................................................6-22 6.5.3 Shielding .........................................................................................................6-22 6.5.4 Geometry ........................................................................................................6-24 6.6 RECORDKEEPING .......................................................................................................6-24 CHAPTER 7 - NUCLEAR CRITICALITY SAFETY 7.0 NUCLEAR CRITICALITY SAFETY .........................................................................................7-1 7.1 REGULATIONS AND STANDARDS ............................................................................7-1 7.2 CRITICALITY CONTROL FACTORS ...........................................................................7-2 7.2.1 Controllable Factors .........................................................................................7-2 7.2.2 Double Contingency Principle ..........................................................................7-4 7.2.3 Administrative Practices ...................................................................................7-5 7.3 CRITICALITY ACCIDENT EXPERIENCE ....................................................................7-6 7.3.1 Types of Criticality Accidents ..........................................................................7-6 7.3.2 Summary of Past Criticality Accidents ..............................................................7-7 7.4 CRITICALITY ALARMS AND NUCLEAR ACCIDENT DOSIMETRY .........................7-8 7.4.1 Criticality Accident Alarm System ....................................................................7-8 7.4.2 Nuclear Accident Dosimetry .............................................................................7-9 7.5 RESPONSIBILITIES OF RADIOLOGICAL CONTROL STAFF .....................................7-11 7.5.1 Routine Operations ..........................................................................................7-12 7.5.2 Emergency Response Actions ...........................................................................7-12 7.5.3 Special Considerations During Decommissioning Activities ...............................7-13 CHAPTER 8 - WASTE MANAGEMENT 8.0 WASTE MANAGEMENT .........................................................................................................8-1 8.1 POTENTIALLY CONTAMINATED WASTES ..............................................................8-1 8.1.1 Solid Waste .....................................................................................................8-1 8.1.2 Liquid Waste ...................................................................................................8-1 8.2 DESIGN OF WASTE PROCESSING SYSTEMS ............................................................8-2 8.2.1 Objectives........................................................................................................8-2 8.2.2 Effluents .........................................................................................................8-2

8.3 TREATMENT ...............................................................................................................8-2 8.3.1 Airborne Waste ...............................................................................................8-3 8.3.2 Liquid Waste ...................................................................................................8-3 8.3.3 Solid Waste .....................................................................................................8-4 viii DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities TABLE OF CONTENTS (continued) 8.4 MONITORING ..............................................................................................................8-4 8.4.1 Air and Gaseous Effluents ................................................................................8-4 8.4.2 Liquid Effluents ...............................................................................................8-4 8.4.3 Water Collection System .................................................................................. 8-5 8.5 WASTE MINIMIZATION ............................................................................................. 8-5 9.0 EMERGENCY MANAGEMENT ............................................................................................... 9-1 9.1 EMERGENCY MANAGEMENT IN DOE ...................................................................... 9-1 9.1.1 Key Emergency Management Principles ........................................................... 9-1 9.1.2 Requirements Pertaining to All DOE Operations ............................................... 9-2 9.2 SPECIFIC GUIDANCE ON EMERGENCY MANAGEMENT FOR URANIUM FACILITIES .......................................................................................................................................................9-2 9.2.1 Hazards Assessment ........................................................................................ 9-3 9.2.2 Program Elements ............................................................................................ 9-5 CHAPTER 10 - DECONTAMINATION AND DECOMMISSIONING 10.0 DECONTAMINATION AND DECOMMISSIONING ..............................................................10-1 10.1 REGULATIONS AND STANDARDS .........................................................................10-1 10.1.1 Other Regulations .........................................................................................10-2 10.1.2 Residual Radioactivity Levels ........................................................................10-3 10.2 DESIGN FEATURES FOR NEW FACILITIES ............................................................10-5 10.2.1 Building Materials ........................................................................................10-5 10.2.2 Ventilation Systems ......................................................................................10-6 10.2.3 Piping Systems .............................................................................................10-6 10.2.4 Soil-Contamination Considerations ................................................................10-7 10.2.5 Other Features ..............................................................................................10-7

10.3 DECONTAMINATION AND DECOMMISSIONING PROGRAM REQUIREMENTS ..................................................................................................10-7 10.3.1 Pre-Operational and Operational Activities .....................................................10-8 10.3.2 Post-Operational Activities ............................................................................10-8 10.3.3 Decontamination and Decommissioning Activities ..........................................10-8 10.3.4 Post-Decommissioning Activities ................................................................10-13 10.3.5 Quality Assurance .......................................................................................10-13 10.4 DECONTAMINATION AND DECOMMISSIONING EXPERIENCE .........................10-13 REFERENCES ........................................................................................................................................................................... R-1 ix DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities TABLE OF CONTENTS (continued) FIGURES 2-1 Specific activity for mixtures of U-238, U-234, and U-235 .................................................... 2-9 2-2 Percent of Total Radioactivity by Isotope vs % Weight U-235 Enrichment ............................. 2-10 2-3 Approximate Percent Alpha Activity Contribution for Laser Enriched Uranium ..................... 2-10 2-4 Estimated Uranium Specific Activity for Laser Enrichment (of natural uranium) .................... 2-11 2-5 U-238 Decay Product Ingrowth ...........................................................................................2-19 4-2 Protocol for Release of Materials ......................................................................................... 4-22 6-1 Beta Radiation Readings at Surface of Uranium Metal vs % Enrichment by Weight ................ 6-5 6-2 Absorbed Dose Rate as a Function of Depth in Mylar ........................................................... 6-6 6-3 Changes in Beta Energy Spectra and Shallow Dose Rate From a Natural Uranium Metal Slab Source Caused by Protective Apparel.......................................... 6-7 6-4 Meter Readings for a Depleted Uranium Ingot ...................................................................... 6-11 6-5 Meter Readings for an Open Drum of UF4 (green salt) .......................................................... 6-12 6-6 Measured Angular Response of the INEL TE Survey Meter to Parallel Beams of Beta Particles From Three Standard Beta Sources .................................................. 6-13 6-7 Average Ion Chamber Survey Meter Response by Group to X or Gamma Photon Radiation ......................................................................................... 6-15 6-8 Average GM Survey Meter Photon Energy Response by Group ............................................. 6-15 6-9 High Resolution Gamma Spectrum of Slightly Enriched Uranium Oxide (1% U-235) recorded with Ge (Li) Detector ............................................................... 6-16 TABLES

2-1 Typical Isotopic Abundances (g of Isotope per 100 g of Material) ........................................... 2-2 2-2 Properties of Radionuclides that May Be Found at Uranium Facilities..................................... 2-3 2-3 Uranium Series (4n + 2) ....................................................................................................... 2-5 2-4 Actinium Series (4n + 3) ...................................................................................................... 2-6 2-5 Uranium Specific Activities ................................................................................................. 2-7 2-6 ALIs and DACs for Uranium and Selected Contaminants in Recycled Uranium ...................... 2-13 2-7 Beta Surface Exposure Rates From Equilibrium Thickness of Uranium Metal and Compounds ....................................................................................................... 2-18 2-8 Toxicological Limits on Airborne Concentrations of Transportable (soluble) Uranium .................................................................................................... 2-22 2-9 Uranium Levels for Various Effects ..................................................................................... 2-23 2-10 1999 ACGIH Threshold Limit Values (TLVs) for Selected Metals ........................................ 2-25 2-11 Inhalation Classification for Some Uranium Compounds ...................................................... 2-26 2-12 Determination of "Dividing Line" Enrichments Above Which Radiological Monitoring Requirements Become Limiting ................................................................................ 2-28 2-13 Impact of Requirement To Monitor at 2%............................................................................. 2-29 4-1 Surface Contamination Values, dpm/100 cm2 ................................................................................................................................................4-30 5-1 Urine Intake Retention Fractions (IRF) and Bioassay Goals for 238U Inhalation of 1-µm AMAD particles .......................................................................... 5-4 5-2 Feces Intake Retention Fractions (IRF) and Bioassay Goals for 238U Inhalation of 1-µm AMAD particles .......................................................................... 5-5 5-3 Lung Intake Retention Fractions (IRF) and Bioassay Goals for 238U Inhalation of 1-µm AMAD particles .......................................................................... 5-6 5-4 Dose Conversion Factors for 238U ......................................................................................... 5-7 x DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities TABLE OF CONTENTS (continued) 5-5 Minimum Uranium Bioassay Monitoring ............................................................................. 5-12 5-6 Categories and Performance Criteria for Uranium Bioassay ................................................... 5-15 5-7 Minimum Suggested Frequencies of Bioassay for Uranium ................................................... 5-16 5-8 GI Tract Absorption Factors for Uranium ............................................................................. 5-23 5-9 Mathematical Model to Describe Clearance from the Respiratory Tract for

the Fisher-Modified Wrenn-Lipsztein Uranium Urinary Excretion Model .................... 5-24 5-10 Fisher-Modified Wrenn-Lipsztein Uranium Model Parameter Values ..................................... 5-25 5-11 ICRP Publication 30 Uranium Urinary Excretion Parameter Values ......................................5-26 5-12 Parameter Values for the Replacement Function for the Fisher-Modified Wrenn-Lipsztein Uranium Model ..............................................................................5-26 5-13 Health Effects from Acute Intake of Soluble Uranium ..........................................................5-29 5-14 Uranium Balance for Reference Man ...................................................................................5-30 5-15 Uranium Levels for Internal Dosimetry Notification .............................................................5-37 5-16 Uranium Contamination Levels for Notification of Occupational Medicine Physician ............5-38 5-17 Early Bioassay Measurement Results Corresponding to the Therapeutic Intervention Action Levels Used at the Hanford Site ......................................................................5-40 6-1 Effective Depth of Tissue for Various Organs ...................................................................... 6-1 6-2 Tissue Weighing Factors .................................................................................................... 6-2 6-3 Radiation Dose Limits for DOE and DOE Contractors ......................................................... 6-3 6-4 Spontaneous Fission Neutron Yields ................................................................................... 6-9 6-5 Neutron Yields from Alpha-Neutron Reactions for Oxides and Fluorides .............................6-10 6-6 Neutron Yields for Trace Impurities in Uranium .................................................................6-10 6-7 Instrument Response to Uranium Beta Fields ......................................................................6-12 6-8 Gamma Flux and Ratios at Various Locations and Sources at Fernald Plant ..........................6-17 6-9 Performance Test Categories, Radiation Sources, and Test Ranges for the DOELAP and NVLAP Programs .............................................................................6-21 6-10 Uranium Beta Shielding ....................................................................................................6-23 6-11 Uranium Beta Dose Reduction Factors ...............................................................................6-23 APPENDICES APPENDIX A - GLOSSARY .......................................................................................................... A-1 x i DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities This page intentionally left blank. x i i DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 1.0 INTRODUCTION 1.1 PURPOSE AND APPLICABILITY This Technical Standard (TS) provides operational guidance, practical lessons learned and experience gained, guides to good practice, and reference information on the safe handling of uranium. The TS provides information to assist uranium facilities in complying with Title 10 of the Code of Federal Regulations (CFR), Part 835, Occupational Radiation Protection (10 CFR 835) (DOE 1998a). This TS supplements the DOE G 441.1 series of Guides, DOE Orders, and DOE-STD-1098-99, Radiological Control (RCS) (DOE 1999a) and has as its sole purpose the protection of workers and the public from the hazards that are inherent in uranium storage, processing, and handling.

This TS replaces EGG-2530, Health Physics Manual of Good Practices for Uranium Facilities (EGG 1988), providing more complete and current information and emphasizing situations that are typical of DOE’s current operations, including weapons assembly and disassembly, safe storage, decontamination, and decommissioning (environmental restoration). This TS may be useful to health physicists and other safety professionals. The information presented herein represents the best technical information available from within the DOE complex. Except to the extent that the guidance presented here is an exact quote from applicable regulations or contract requirements, it is not binding or mandatory. However, judicious use of this TS, in concert with applicable regulatory documents, will help in building a comprehensive and technically-defensible radiological control program. Regulatory guidance and references are current as of October 2004. 1.2 DEFINITIONS A glossary is provided in Appendix A. In all cases, the definitions provided in this TS are consistent with those provided in 10 CFR 835, its Guides, and the RCS. 1.3 DISCUSSION Chapters 2 through 10 provide technical information to assist in safely managing radiological hazards associated with uranium operations. The topics covered are those considered by representatives of many of DOE’s uranium facilities to be most beneficial: Properties and Relative Hazards (Chapter 2), Radiation Protection (Chapter 3), Contamination Control (Chapter 4), Internal Dosimetry (Chapter 5), External Dose Control (Chapter 6), Nuclear Criticality Safety (Chapter 7), Waste Management (Chapter 8), Emergency Management (Chapter 9), and Decontamination and Decommissioning (Chapter 10). 1- 1 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities This page intentionally left blank. 1- 2 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 2.0 PROPERTIES AND RELATIVE HAZARDS This chapter presents basic radiological and chemical properties of uranium and discusses the basis for current control limits. A variety of materials are inherent to uranium handling processes and hazards characteristic of these materials and processes. The data and discussions are intended to provide a basis for understanding the changes in hazards as a function of such parameters as enrichment, physical form, and chemical form. 2.1 NUCLEAR PROPERTIES OF URANIUM Naturally occurring uranium consists of a mixture of 234U, 235U and 238U isotopes, along with their decay products. Uranium is relatively abundant in nature. The primary isotopes of uranium are long- lived alpha-emitters with energies between 4.15 and 4.8 MeV. Their progeny include numerous other radionuclides, some of which are radiologically significant at uranium facilities, the degree of significance depending upon the history of the uranium materials and the processing. Through proper processing, uranium can be used as a fuel in nuclear reactors to generate electricity on a commercially-viable scale. The 235U isotope is readily fissioned by slow, "thermal" neutrons with the release of a large amount of energy. The percentage of 235U present (referred to as "enrichment") determines the fuel reactivity and the criticality hazard of the material. By concentrating the amount of the 235U isotope in the uranium, the quantity of fuel and the size of the reactor needed for production decreases. This concentration of natural uranium to enriched uranium is carried out by special processes such as gaseous diffusion, centrifuging, or laser separation. The uranium by-product of the enrichment process is reduced in 235U content and is called "depleted" uranium. Uranium is commonly classified by its 235U enrichment as natural uranium, enriched uranium, or depleted uranium.

Uranium-235 fissions after capturing a thermal (very low energy) neutron. Its fission thermal cross- section (probability of interaction) is 577 barns (Stehn et al. 1965). Its neutron capture cross section is 101 barns. After capturing a fast neutron, 238U undergoes two successive beta decays to 239Pu which will also undergo thermal fission (thermal cross-section = 741 barns). Pressurized heavy-water reactors function with natural uranium isotopic composition. Other types of reactors require some 235U enrichment. 2.1.1 Isotopic Characterization Natural uranium consists of three isotopes: 238U, 235U, and 234U. All three radionuclides undergo radioactive decay by alpha particle emission. The 235U isotope (and 234U to a much lesser degree and at lower energy) emits gamma radiation as well. The natural abundances of these uranium isotopes, as well as the weight percentages of the isotopes in enriched (typical commercial nuclear power reactor enrichment) and depleted uranium, are listed in Table 2-1. 2- 1 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Table 2-1. Typical Isotopic Abundances Isotope Natural Typical Commercial Feed Enrichment Depleted Specific Activity Ci/g Neutron Capture Cross Section (barns) 238U 99.28 97.01 99.80 3.3 E-7 2.7 235U 0.72* 2.96 0.20 2.1 E-6 101 234U 0.0055 0.03 0.0007 6.2 E-3 95 The amount of uranium present determines the grade of the ore. Most of the ores found in the U.S. contain from 0.1 to 1% uranium and are considered medium grade. Lower-grade ores are mined commercially if they are a byproduct of mining for another material, such as gold or phosphate. Uranium that has been processed to raise the concentration of 235U is referred to as enriched uranium. The extent of enrichment depends on the intended end use of the uranium. Commercial light water reactors are designed for use with the 235U enriched to around 3%. Higher enrichment is required for; high-temperature gas-cooled reactors, naval nuclear propulsion reactors, most research reactors and weapons. The 235U enrichment process also increases the concentration of 234U. The higher activity of enriched uranium is due more to increased 234U than to increased 235U. Depleted uranium is a by-product of the enrichment process and is depleted in both the 235U and 234U isotopes. Depleted uranium, with its reduced activity and very high density, has many uses; among them are radiation shielding, counterweights, projectiles, and target elements in DOE plutonium production reactors. In addition to the uranium isotopes discussed above, the daughter products of uranium decay and byproducts of uranium processing can have significant radiological impacts in uranium-handling facilities. Table 2-2 presents the properties of these radionuclides. 2- 2 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities Table 2-2. Properties of Radionuclides That May Be Found at Uranium Facilities(a) Energies (MeV) and Abundances of Major Radiations (All low yield radiations are not included) Nuclide Half-Life Alpha Beta Gamma Primary Uranium Isotopes 238U 4.51 x 109 y 4.15 (21%) 4.20 (79%) 235U 7.1 x 108 y 4.21 (6%) 4.37 (17%) 4.40 (55%) 4.60 (5%) 0.144 (11%) 0.163 (5%) 0.186 (57%) 0.205 (5%) 234U 2.47 x 105 y 4.72 (28%) 4.77 (72%) 0.053 (0.12%) Decay Products 234Th 24.1 d 0.103 (21%) 0.193 (79%) 0.013 (9.8%) 0.063 (3.5%) 0.092 (3%) 0.093 (4%)

234mPa 1.17 m 2.29 (98%) 0.765 (0.30%) 1.001 (0.60%) 231Th 25.5 h 0.206 (13%) 0.287 (12%) 0.288 (37%) 0.305 (35%) 0.026 (2%) 0.084 (10%) Impurities (e.g. irradiation and reprocessing artifacts) 99Tc 2.12x105 y 0.292 237Np 2.14x106 y 4.78 (75%) 4.65 (12%) 238Pu 86.4 y 5.50 (72%) 5.46 (28%) 239Pu 2.44x104 y 5.16 (88%) 5.11 (11%) 240Pu 6.6x103 y 5.17 (76%) 5.12 (24%) 241Pu 13.2 y 0.021 232U 72 y 5.26 (31%) 5.32 (69%) 236U 2.34x107 y 4.47 (24%) 4.52 (76%) (a) From EGG-2530 (1988). 2- 3 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities 2.1.2 Decay Chains The natural uranium isotopes decay by alpha emission. The decay products are also radioactive and form "decay chains" that ultimately lead to a stable isotope of lead. Tables 2-3 and 2-4 present the decay chains of 238U and 235U (234U is a member of the 238U decay chain), along with the half-lives and characteristic radiations of each nuclide. Uranium-processing steps (milling or refining) separate the decay products and other impurities in the ore from the uranium. It takes months after processing before the first few decay products build up and come to equilibrium with the parents. In depleted uranium, the beta radia tion from the decay of 234Th and 234mPa amounts to nearly twice the alpha radiation from 238U and 234U. In commercially enriched uranium, the beta radiation from 231Th, 234Th, and 234mPA nearly equals the alpha radiation from 238U, 234U, and 235U. In natural ore, the later decay products (especially 230Th and 226Ra) are present and add significant gamma radiation to the emitted radiation. In processed uranium (natural, enriched, or depleted) all decay products below 234U and 235U are removed. Because of the long half-lives of 234U and 231Pa the radionuclides that follow these two nuclides are generally ignored. The mining and milling stages are usually conducted by commercial enterprises. DOE facilities do not routinely process uranium ore concentrates and, as a result, the decay products formed during DOE processing operations of virgin feed are limited. However, radium and its progeny may be present in waste water streams of certain facilities, so it is prudent to consider those nuclides in effluent and environmental monitoring programs. For workplace radiological controls, 234Th, 234mPa, 231Th and the uranium isotopes are those requiring primary consideration; however, if there are large quantities of aged highly enriched uranium, there may be a need to also consider 231Pa in establishing radiological controls. In addition, elevated radon concentrations can occur in poorly ventilated uranium storage areas from the small amounts of 226Ra that grow in and carry over as contaminants in the chemical separation processes. 2- 4 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities 2-5 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities 2-6 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities 2.1.3 Enrichment

Uranium-235 enrichment processes selectively increase the 235U concentration by separating it from the 238U. The method used for many years in the U.S. is the gaseous diffusion process. Laser separation has also been demonstrated in this country, but a facility built to accommodate the process has not yet been brought on-line. Centrifugation is a third separation method used by foreign sources. Uranium feed for the enrichment process is derived from virgin ore or from "very clean" recycled material. Although some uranium is still mill-derived, much of the feed is recycled material from other countries, including Canada (where natural uranium is the reactor feed material). Specifications on acceptable contamination levels limit the feed that may be processed in the U.S. gaseous diffusion plants. Recycling of reprocessed (irradiated uranium) material from DOE’s reactors years ago contaminated the diffusion process equipment with transuranics, a portion of which remains in the equipment. The specific activity of essentially pure uranium depends on its degree of enrichment and normally describes only alpha activity. The beta activity from associated decay products is not included in the uranium specific activity values, but is expressed separately. Consequently, two specific activities (one for alpha and one for beta) are frequently calculated for uranium-bearing materials. Some typical alpha specific activity va lues are given in Table 2-5 and Figures 2-1, 2-2, 2-3, and 2-4. Table 2-5. Uranium Specific Activities Type Wt. % 235U Specific Activity (Ci/g) of Mixture Natural 0.72 7 x 10-7 Depleted 0.20 4 x 10-7 Enriched 2.0 1 x 10-6 Enriched 20 9 x 10-6 For gaseous diffusion enriched uranium, the approximate alpha specific activity of a given uranium enrichment can be calculated from the following formula: Specific Activity of Enriched Uranium = (0.4 + 0.38E + 0.0034E2) x 10-6 Ci/g where E = % 235U by weight, enrichment > or = 0.72 Gaseous diffusion, the predominant existing enrichment technology, causes a greater increase in 234U concentration than in 235U concentration. For example, when 235U content is increased from 0.72% (natural) to 2.96%, (an increase of approximately a factor of four), 234U content increases from 0.006% to 0.03%, (a five-fold increase). As a result, the specific activity increases with enrichment, not just because of the replacement of some 238U with 235U, but more significantly because of the increase in the amount of 234U present. Laser isotopic separation (under research) selectively enriches only the 235U, leaving the 234U with the "tails," or depleted uranium. Therefore, the radiological characteristics of both enriched and depleted uranium will change when compared to conventional separation techniques. Figures 2-3 and 2- 4 illustrate this effect. The specific activity of recycled irradiated uranium varies from the value calculated from the equation given above because that equation is not applicable to recycled material with its added contaminants. For these reasons, specific activities that are calculated from the formula should be 2- 7 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities considered approximations only. If exact values of specific activity are required, they should be determined analytically. See Example 1 for the calculation of blending enrichments. 2-8 Example 1 One kilogram of 20% enriched uranium is blended with 1 kilogram of 2% enriched uranium.

SA = [0.4 + 0.38E + 0.0034E2] x 10-6 Ci/g SA20 = [0.4 + 0.38 (20) + 0.0034 (20)2] x 10-6 Ci/g = 9.36 x 10-6 Ci/g SA2 = [0.4 + 0.38 (2) + 0.0034 (2)2] x 10-6 Ci/g = 1.17 x 10-6 Ci/g The specific activity of the resulting mixture is [(9.36 + 1.17)/2] x 10-6 Ci/g = 5 x 10-6 Ci/g DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities Figure 2-1. Specific Activity for Mixtures of 238U, 234U, and 235U 2-9 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities Figure 2-2. % Total Radioactivity by Isotope vs. % Weight 235U Enrichment Calculated from SA = (0.4 + 0.38E+0.0034E2) 10-6 Ci/g (gaseous diffusion process) (NRC Reg Guide 8.11) Figure 2-3. Approximate Percent Alpha Activity Contribution for Laser Enriched Uranium 2-10 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities Figure 2-4. Estimated Uranium Specific Activity for Laser Enrichment (of natural uranium) The Derived Air Concentration (DAC) values for several radionuclides are shown in Table 2-6. Because the ALIs for the three primary uranium isotopes are expressed in activity units, enrichment has little impact on inhalation and ingestion ALIs. However, as illustrated in Table 2-5, as enrichment increases from 2% to 20%, the specific activity increases nine-fold. Consequently, the mass of material that corresponds to one ALI decreases by a factor of nine. The degree of enrichment also affects the controls that are required for external penetrating radiation exposure because of the increase in the amount of gamma-emitting 235U that is present. As a historical note, some of the earlier documentation refers to the "special curie" of natural uranium, which was defined as 3.7 x 1010 d/s of 234U, 3.7 x 1010 d/s of 235U, and 1.7 x 109 d/s of 235U. Thus, 1 "curie" of natural uranium was actually slightly more than 2 curies of uranium alpha activity. This essentially obsolete term has caused considerable confusion. Readers are cautioned to be aware of the use of this special curie in the older literature. Use of this unit in any current application is strongly discouraged. 2.1.4 Contaminants from Recycled Uranium and Associated Hazards Some of the uranium feed material that was handled at DOE facilities had been reclaimed or recycled from reprocessed, spent reactor fuel. The chemical processes by which recycled uranium was purified left trace amounts of transuranic elements (neptunium, americium, and plutonium) and fission products (mainly 99Tc). The recycled uranium also contained trace amounts of uranium isotopes not found in nature, such as 236U. At the minute concentration levels in uranium from fuel reprocessing facilities, the radiological impact of these impurities was negligible in most cases. However, there were many routine chemical processes that tended to concentrate these impurities, either in the uranium product or in reaction by-products, such that radiological controls and environmental monitoring programs must consider these impurities. 2-11 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities The following text discusses the environmental, safety, and health challenges presented by the introduction of recycled uranium into the DOE system for enrichment. 2.1.4.1 Transuranics

Transuranics (neptunium and plutonium isotopes) exist in small quantities in reclaimed or recycled feed materials. In most cases, a regimen of radiological controls based on uranium hazards is adequate to control the additional activity. However, because of their higher specific activities (compared to uranium isotopes), transuranics can represent a significant internal dose concern even at very low mass concentrations. As a result, the ALIs for transuranics are lower than those for uranium isotopes. For example, for a moderately soluble transportability mixture, if 239Pu contamination contributes 0.1% of the total alpha activity in uranium, then it will contribute roughly 14% of the total inhalation dose equivalent (see Example 2). Example 2 illustrates that it takes only 11 parts of 239Pu per billion parts of natural uranium to attain an activity fraction of 0.1%. Radiological controls based solely on uranium content may provide insufficient protection with increases in the TRU concentration. Processes to recover uranium from by-product streams recover a portion of the impurities as well and may require additional controls to adequately protect individuals when the TRU concentration exceeds 0.1%. 2-12 DOE-STD-1136-2004 _ Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities Table 2-6. ALIs and DACs for Uranium and Selected Contaminants in Recycled Uranium Inhalation Nuclide Class D * Class W * Class Y * Annual Limits on Intake (Bq values converted from the rounded off ALIs), from 10 CFR 20 uCi (Bq) uCi (Bq) uCi (Bq) 238U 235U 234U 234Th 234mPa 231Th 99Tc 237Np 238Pu 239Pu 240Pu 241Pu 236U 1 (4 x 104) 1 (4 x 104) 1 (4 x 104) NL1 NL NL 5 x 103 (2 x 108) NL NL NL NL NL 1 (4 x 104) 8 x 10-1 (3 x 104) 8 x 10-1 (3 x 104) 7 x 10-1 (3 x 104) 2 x 102 (7 x 106) NL 6 x 103 (2 x 108) 7 x 102 (3 x 107) 4 x 10-3 (1 x 102) 7 x 10-3 (3 x 102) 6 x 10-3 (2 x 102) 6 x 10-3 (2 x 102) 3 x 10-1 (1 x 104) 8 x 10-1 (3 x 104) 4 x 10-2 (1 x 103) 4 x 10-2 (1 x 103) 4 x 10-2 (1 x 103) 2 x 102 (7 x 106) NL 6 x 103 (2 x 108) NL NL 2 x 10-2 (7 x 102) 2 x 10-2 (7 x 102) 2 x 10-2 (7 x 102) 8 x 10-1 (3 x 104) 4 x 10-2 (1 x 103) Inhalation DAC, From 10 CFR 835, Appendix A uCi/mL (Bq/m3) uCi/mL (Bq/m3) uCi/mL (Bq/m3) 238U 235U 234U 234Th 234mPa 231Th 99Tc 237Np 238Pu 239Pu 240Pu 241Pu 236U 6 x 10-10 (2 x 101) 6 x 10-10 (2 x 101) 5 x 10-10 (2 x 101) NL NL NL 2 x 10-6 (8 x 104) NL NL NL NL NL 6 x 10-10 (2 x 101) 3 x 10-10 (1 x 101) 3 x 10-10 (1 x 101) 3 x 10-10 (1 x 101) 9 x 10-8 (3 x 103) 3 x 10-6 (1 x 105) 3 x 10-6 (1 x 105) 3 x 10-7 (1 x 104) 2 x 10-12 (9 x 10-2) 3 x 10-12 (9 x 10-2) 2 x 10-12 (8 x 10-2) 2 x 10-12 (8 x 10-2) 1 x 10-10 (4) 3 x 10-10 (1 x 101) 2 x 10-11 (6 x 10-1) 2 x 10-11 (6 x 10-1) 2 x 10-11 (6 x 10-1) 6 x 10-8 (2 x 103) 3 x 10-6 (1 x 105) 3 x 10-6 (1 x 105) NL NL 7 x 10-12 (3 x 10-1) 6 x 10-12 (2 x 10-1) 6 x 10-12 (2 x 10-1) 3 x 10-10 (1 x 101) 2 x 10-11 (6 x 10-1) 2- 13 1 NL = Not listed. * See last paragraph of Section 2.5 for discussion of Class D, W and Y. DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 2-14 Example 2 One gram of natural uranium contains 239 Pu contamination to the extent that the 239 Pu activity is 0.1% of the uranium alpha activity. The relative inhalation hazards of the two materials are determined by dividing each material’s relative activity by its derived air concentration.

U-Nat relative activity = 1 239 Pu relative activity = 0.001 U-Nat derived air concentration (W) = 3 x 10-10 µCi/mL 239 Pu derived air concentration (W) = 2 x 10-12 µCi/mL These values represent the relative hazards of the two materials in the mixture. Therefore, 239 Pu at 0.1% of the U-Nat activity represents 14% of the potential inhalation dose. The activity of 1 gram of U-Nat = 2.5 x 104 dps Therefore, 0.001 x 2.5 x 104 = 2.5 x 101 dps = the 239 Pu activity in the 1 gram of U-Nat. The specific activity of 239 Pu is 2.27 dps/nanogram: Therefore, 0.1% 239 Pu activity fraction corresponds to 11 parts per billion on a mass basis. DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities Several DOE facilities have adopted specifications on recycled uranium that limit the amount of transuranic alpha activity to 0.1% of the total uranium alpha activity, thus limiting the potential inhalation dose from transuranics to a small fraction of the total. Facilities that handle recycled uranium with higher levels of transuranics should establish a regular program of analyzing feeds, products, and by-products for transuranics, and then modifying control limits and action levels as appropriate to reflect the transuranic content of those materials. This monitoring of the TRU content is essential when the analytical technique used to identify the level of radiological control needed is based on gross alpha counting (such as for air sampling), which does not distinguish the plutonium from the uranium fraction, or chemical analysis for uranium (such as photofluorometric urinalysis) which does not detect plutonium. Raffinate from refinery operations, MgF2 from metal production operations, and chemical traps from UF6 operations have all been observed to have higher TRU-to-U ratios than either reactants/feeds or uranium products. Frequently, reaction by-products are not discarded as wastes but are processed further to recover the remaining uranium. When this occurs, a portion of the impurities is recovered along with the uranium and can become a perpetual radiological control problem. All facilities that process recycled uranium should periodically analyze feeds, products, and by-products for transuranics to ensure that radiological controls are adequate for the mixtures of uranium and transuranic elements that are present. The uranium isotopes (viewed as contaminants) that will increase due to the recycled uranium feed are 232U, 234U, and 236U. The health and safety risks of 236U are similar to those of natural uranium because its specific activity and radiation emissions are similar (See Table 2-2). Its presence in uranium fuel requires slightly higher enrichments for the same reactor applications, however, because it absorbs neutrons. The increased concentration of the 234U increases the specific activity of any enrichment of 235U. It is expected that the specific activity for a given enrichment would be about double that obtained from enrichment of non-recycled uranium. The isotope in recycled uranium presenting the greatest potential radiological hazard from external

sources is 232U. 232U a daughter product of neutron activation of 231Pa. The health hazards of 232U are primarily due to the rapid buildup of gamma activity of its decay products, particularly from 228Th. The gamma activity buildup is both time- and process-dependent. The 232U decay products form nonvolatile fluorides and will concentrate in cylinders when UF6 is vapor-fed. The gamma activity in equipment that processes gaseous UF6 is a function of the mass fraction of 232U present in the gas phase. Estimates indicate that the level of gamma activity within the enrichment cascade equipment would increase by about a factor of 3 due to the presence of 232U. The exposure rates on internal surfaces would increase from 10-20 mrad/h to 30-60 mrad/h; those on external surfaces would increase to about 3-4 mrad/h. The major exposure increase from the 232U occurs in the handling of UF6 cylinders. Currently, the exposure rate at the external surface of empty UF6 cylinders is about 50-100 mrad/h. Assuming a 232U concentration of 0.5 ppm based on 235U and a feed enrichment of 1%, a full 10-ton feed cylinder would have a surface exposure rate of about 80 mrad/h. The exposure rate at 30 cm from the surface of an emptied cylinder would be about 500 mrad/h without the shielding provided by material in the cylinder. These values are based on the 232U being in secular equilibrium with its decay products; in reality, it is unlikely that the decay products would reach much more than 50% of equilibrium values. Product cylinders produced from processing of recycled uranium typically have higher gamma radiation fields than the feed cylinders. At 4% 235U enrichment, the contribution from 232U over time could increase the radiation field at the surface from 80 mrad/h to 300 mrad/h from a full 10-ton cylinder and from 500 mrad/h at 30 cm to 2 rad/h from an empty cylinder. About half of this increase would be 2-15 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities apparent within 2 years of initial usage and the highest levels could occur in 20 years without mitigating actions. Frequent cylinder cleaning can prevent this significant exposure rate buildup. The presence of 232U may also require other changes in processes used to handle cleaning solutions due to the higher gamma radiation present. 2.1.4.2 Technetium In facilities with significant quantities of 99Tc, radiation monitoring techniques must be able to detect the low-energy beta radiation from this isotope. Individual and area monitoring equipment and techniques selected to measure the 2.29 MeV (Emax) beta from 234mPa may not measure the 99Tc 0.292 MeV (Emax) beta effectively. If a mixture of uranium and 99Tc is suspected to be present, the monitoring technique selected must be based on 99Tc or on the actual mixture, rather than on 234mPa. The 99Tc levels have not been the controlling factor in many situations to date. However, it is important to ensure that monitoring instruments and techniques are adequate to detect 99Tc.

Technetium-99 tends to deposit within enrichment equipment and will "pocket" in the higher enrichment sections of the gaseous diffusion cascade. Special precautions must be taken when evacuating and purging or performing other maintenance work on this equipment. In equipment with accumulations of 99Tc, low energy beta radiation fields of a few rad per hour may be encountered. This radiation is effectively attenuated by the protective clothing required for contamination control (one pair of industrial cloth coveralls, one pair of impermeable (Tyvek) coveralls and heavy neoprene gloves). While the 99Tc should be effectively removed from the Gaseous Diffusion Plant (GDP) product, it will be present in uranium used by other DOE facilities. Because the ALI for 99Tc is higher than that of uranium, inhalation is the controlling concern only in situations where the technetium activity greatly exceeds that of the uranium that is present. Technetium as pertechnetate is also difficult to remove from skin and can therefore cause significant skin doses from skin contamination. The tendency of technetium to become airborne more readily than uranium can lead to beta contamination in areas where it is not otherwise expected and environmental emissions even when the uranium is effectively confined in the work place. Residues in ventilation systems from high-temperature operations, such as uranium remelting/casting, or uranium chip burning, tend to have higher Tc-to-U ratios than either feed or product material in uranium metal processing facilities. Because of its low atomic weight and relative volatility, technetium also tends to concentrate at the top of the gaseous diffusion cascade, where it becomes an inhalation and effluent concern when the cascade is opened for maintenance. Facilities that handle recycled uranium should 1) analyze feeds, products, and by-products to determine the fate of 99Tc within their processes, then 2) modify monitoring equipment, control limits, and action levels as needed to properly evaluate and control 99Tc hazards. Environment, safety and heath personnel should also evaluate the presence of and radiological consequences from other fission products impurities in recycled uranium. 2.2 PHYSICAL AND CHEMICAL PROPERTIES Uranium fuels vary with reactor type. Some reactors use the natural isotopic composition in the fuel. Others use enrichment varying from 2% to > 90%. Because of the radiation-induced growth of uranium metal used in the early reactors, alloys were developed to stabilize dimensional changes. Many of the alloys with favorable dimensional stability characteristics had sizeable neutron absorption cross-sections, resulting in poisoning of the nuclear reaction. Zirconium-clad ceramic uranium dioxide and uranium carbide fuels were found to have acceptable characteristics and are in common use. 2-16 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities 2.2.1 Uranium Fuel Processing

The process of reducing uranium ore to metal begins with the discovery and mining of uranium in ore bodies. Most medium grade ore consists of oxides of uranium, of which carnotite (K2(UO2)2(VO2.3H2O)) is predominant. Although some ore is mined using in situ leach techniques, most is hard-rock mined with a small amount removed by open pit mining. Uranium ore is milled by crushing, leaching, extracting, and precipitating, usually to ammonium diuranate ((NH4)2U2O7) commonly called yellow cake. The radioactivity of this product is low because the decay products have been stripped away and it is in an unenriched form. The yellow cake is purified and converted to UF4 and then further fluorinated to uranium hexafluoride (UF6). Gaseous diffusion enrichment changes the uranium isotopic, but not the chemical, composition of the gas. The UF6 is hydrolyzed to uranyl oxyfluoride, which is precipitated with an ammonia solution to ammonium diuranate. This precipitate is filtered or centrifuged, dried, and calcined. The uranium compound is reduced to UO2 powder, which is pelletized, sintered, and encapsulated in tubes for reactor usage. Laser enrichment can use feed forms including metal and UF6. Steel was an early cladding material that was discontinued because of its thermal-neutron poison characteristics. Fuel bundles used in commercial LWRs are now made of fuel pins that consist of pellets of UO2. The pellets are stacked into free-standing cladding tubes of a zirconium or zirconium-tin alloy. Differences in fuel design between the two common types of nuclear reactors in use in the United States, pressurized water reactors (PWRs) and boiling water reactors (BWRs), are rod diameter and cladding thickness. Reactor fuel for the Canadian pressurized heavy water reactors (CANDU-PHWR) is similar but the cladding need not be free-standing. Additionally, the fuel pins are smaller in diameter. Breeder reactors like the Fast Flux Test Facility (FFTF) use a mixture of PuO2 and depleted UO2. In the case of the FFTF, the pelle ts are loaded into stainless steel cladding tubes (which have a smaller effect on fast neutrons). Uranium carbide (UC2) microspheres were developed as an alternative to UO2, primarily for the high-temperature gas-cooled reactor. These fuel particles, developed for high thermal and radiation stability, prevent the release of fuel and fission products over a wide range of conditions. 2.2.2 Uranium Metal Conversion of UF6 to uranium metal involves, first, the production of UF4, commonly called green salt. Enriched uranium green salt is reacted with granular calcium to produce metal slag. This product is then reacted with magnesium or calcium to reduce the material to metal. Depleted uranium green salt is more commonly reacted with magnesium to produce DU metal as a derby. In both cases, most of the uranium decay products are concentrated in the calcium or magnesium slag, leaving the metal relatively pure and with a reduced level of radioactivity. Buildup of decay products to near-equilibrium levels takes about six months. The metallic uranium is processed into desired forms using machining, melting, casting, and other treatments. This very dense metal is usually alloyed with another metal for greater stability. Uranium is a reactive metal that oxidizes easily. In the newly minted metal, a very thin surface layer tends to undergo rapid oxidation. This surface layer may protect the rest of the metal from further corrosion, and prevent the generation of removable contamination. Certain environmental conditions, particularly moist

2-17 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities air and saline solutions, can accelerate the corrosion of the material over time and produce greater possibility for generating airborne radioactive material. Stored in a dry environment or coated with an anti-corrosion surface treatment, the metal may show no visible signs of corrosion for many years. Uranium metal may be dissolved using nitric acid, which is also used to passivate ("p ickle") the metal to inhibit oxidation. 2.3 RADIOLOGICAL CHARACTERISTICS AND EFFECTS Uranium isotopes decay by alpha particle emission and some also emit low-energy gamma rays. For Classes W and Y material (See last paragraph of Section 2.5 for discussion of Class D, W and Y), the inhalation hazard from alpha particle release in the respiratory tract is the predominant radiological hazard associated with the alpha-emitting uranium isotopes. The primary uranium decay products, listed in Table 2- 2, decay by beta particle emission, most with a small yield of gamma emissions as well. These decay products increase the shallow dose equivalent and lens of the eye dose equivalent resulting from external radiation exposures, due mainly to the 2.29 MeV (Emax) beta from 234mPa. The surface exposure rates shown in Table 2-7 result primarily from beta radiation from decay products. The exposure rates decrease quickly with distance because of the attenuation of the beta radiation and the small yield of the gamma radiation. Table 2-7. Beta Surface Exposure Rates from Equilibrium Thickness of Uranium Metal and Compounds Source Beta Surface Exposure Rate, mrad/h U-Nat metal slab 233 UO2 207 UF4 179 UO2(NO3)26H20 111 UO3 204 U3O8 203 UO2F2 176 Na2U2O7 167 Beta surface exposure rate in air through a polystyrene filter 7 mg/cm2 thick. Because some uranium decay products have short half-lives (on the order of days), those decay products will usually be present with uranium during processing. Figure 2-5 illustrates the ingrowth of the 238U decay products. An assumption of secular equilibrium should not be made until processing is complete because many routine chemical processing steps separate uranium from its decay products. Both the inhalation and external exposure hazards associated with the decay products are increased in areas where the decay products are concentrated. The overall inhalation hazard will typically decrease in those areas as the uranium is removed. In the case of cast uranium metal, the exposure rates from high 2-18 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities beta levels from decay products may be many orders of magnitude greater than the exposure rates from the uranium. Figure 2-5. 238U Decay Product Ingrowth 2.3.1 Alpha-Neutron External Hazard The interaction of alpha particles from uranium with the nuclei of fluorine and other low-Z atoms generates neutrons of approximately 2-MeV energy. The magnitude of the neutron flux varies, based on the total activity of uranium (which is a function of enrichment) and the chemical compound in question (mixing of U and F). In the case of UF6, the typically measured neutron dose rates for cooled storage cylinders are as follows: Natural-5% enrichment: 0.01-0.2 mrem/h Very high enrichment (97+%) : 2-4 mrem/h (contact) 1-2 mrem/h (3 ft)

The preceding values were measured with a 9-in. spherical BF3 rem meter. In general, the exposure potential of personnel to neutrons generated by the (alpha, n) reaction is not high. However, if personnel are required to spend more than a few hours per week in close proximity to containers of uranium fluoride compounds or if their assignments require them to spend time near storage or processing areas for large quantities of uranium fluoride compounds, the exposure to neutrons should be evaluated. This is particularly necessary since the personnel monitoring badges may not be neutron-sensitive or may need to be calibrated to the specific spectra. Penetrating radiation exposures from 2-19 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities photon radiation will not be indicative of neutron exposures. This is because the higher photon penetrating radiation exposures tend to be associated with used but empty containers, where decay products have plated out on the sides, while the maximum neutron exposures are associated with full containers. There is a small additional neutron flux from spontaneous fission associated with full containers. Neutron sensitive personnel monitoring badges are recommended for operations dealing with uranium fluoride compounds. 2.3.2 Mode of Uranium Entry into the Body Work practices are designed to control radiation exposure to levels that are as low as is reasonably achievable (ALARA). Reductions in exposure time and increases in shielding help reduce external doses. Effective contamination control techniques and ventilation/filtering systems help reduce airborne radioactive material concentrations and resulting internal doses. Where complete contamination control is not reasonable, internal exposure of uranium compounds as aerosols or deposited particulates may occur. The effects of uranium exposure on the body depend on the mode of exposure. External exposure concerns are limited to beta and gamma emissions, of which the gamma field is quite low and the beta field may be mitigated using protective clothing including safety glasses with side shields. Internal exposure and its potential effects through radiological or chemical toxicity depend on the route of entry, and its distribution depends on the solubility of the material. Solubility is complicated by the wide variety of stoichiometric and crystalline uranium compounds. Inhalation and ingestion are most commonly assessed as routes of entry. Although not covered here, entry of uranium into wounds is also a concern, and its distribution depends on its solubility (See sections 5.9 and 5.10 for further discussion). Absorption through intact skin is unlikely. The type of radiation to which the body is exposed and the length of the exposure determine the biological effect of the radiation exposure. 2.3.2.1 Inhalation

Inhalation hazards from uranium result primarily from the alpha emissions. Inhalation of uranium particles and deposition into the respiratory system are dependent on particle size. The nasal-pharynx system filters out most large particles that are still small enough to be inhaled. Larger particles can be inhaled--a common convention is to assume inhalation possible for all particles 10-µm or less aerodynamic equivalent diameter (AED)--but most particles that penetrate to the lower respiratory tract are less than 3- or 4-µm AED. Uranium in the lungs has been shown to exhibit a wide range of retention values. Clearance may occur through physical processes removing particles that are not embedded into the lung by cilia motion to the esophagus. Uranium particles that are soluble in lung fluid are chemically dissolved, and the ions are transported into the bloodstream where they are further distributed. Uranium particles remaining in the lung constitute a potential radiological hazard as they impart their alpha emission energy into the surrounding absorbing tissue, potentially causing significant damage within a small sphere around each particle. Particles removed from the lung to the bloodstream primarily represent a potential chemical hazard. The significance of these hazards is evaluated using models of uptake and removal recommended by national and international scientific radiation protection organizations. The lung model described in ICRP Publication 66 (ICRP 1994) uses solubility Types of F (fast), M (moderate), and S (slow). In comparison to previous models, this model better describes deposition, retention, and clearance data and decouples physical and chemical clearance processes. 2-20 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiation Protection in Uranium Facilities 2.3.2.2 Ingestion Appropriate uranium contamination controls should prevent ingestion of uranium. Nevertheless, the potential exits for accidental ingestion of uranium. Particles inhaled through the mouth and temporarily deposited there are removed from the respiratory system to the esophagus. Deposition and removal of ingested uranium are approximated using the Gastrointestinal (GI) Tract Model adapted from Eve (Eve 1966). This model calculates material transferred from the GI tract to the blood based on solubility classes (ICRP 1979 and IAEA 1994) or based on a single value for all compounds, as described in ICRP Publication 69 (ICRP 1995). Distribution of uranium transferred into the bloodstream is calculated using a once-through metabolic model. ICRP Publication 30 also provides values for this distribution and excretion to calculate committed doses and long-term tissue retention. Recent models (Wrenn et al. 1994 and ICRP 1995) have been developed to include recycling of uranium back into the blood. 2.4. CHEMICAL TOXICITY The chemical toxicity of uranium is a primary concern in establishing control limits. A heavy metal, uranium is chemically toxic to kidneys and exposure to soluble (transportable) compounds can result in renal injury. The factors to be considered in determining whether the chemical or radiological hazard is controlling are the enrichment, mode of entry, and the solubility/transportability of the material. Chemical toxicity is a higher risk with soluble material of 10% or less enrichment.

A concentration of 3 µg of uranium per gram (µg U/g) of kidney tissue has traditionally been used as the guideline for controlling the chemical toxicity of uranium. Reference man has a kidney mass of 310 g, so this concentration translates to a total kidney burden of 1 mg. A review of the literature by Leggett (Leggett 1989) suggests that worker exposure to 2 to 6 µg U/g kidney might be tolerated with no serious effects. However, he emphasizes that this range is not necessarily the same as the level causing no detectable damage. He concludes that a lower limit would be prudent until more of the physiological mechanisms of response to uranium in the kidney are better understood. Other studies (McGuire 1991) report that detectable effects from an intake of soluble uranium of 10 mg or less is unlikely and that an intake of 40 mg and perhaps as high as 100 mg is unlikely to cause permanent damage. Other evaluations of toxicity to the kidney concluded that a limit of 1.0 µg U/g kidney is consistent with results in the recent literature. An airborne concentration limit of 0.2 mg/m3 was adopted by the Nuclear Regulatory Commission (NRC) and the American Conference of Governmental Industrial Hygienists (ACGIH) for occupational exposures, based on the 3 µg/gm of tissue value. The Occupational Safety and Health Administration (OSHA) has adopted a limit of 0.05 mg/m3 for soluble uranium and 0.25 mg/m3 for insoluble uranium. In most DOE facilities, the most conservative of the two standards (OSHA or ACGIH) should be used unless enrichment and solubility dictate more stringent controls based on radiological concerns. Table 2-8 lists airborne concentration limits for transportable uranium that have been published by various organizations. 2-21 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Table 2-8. Toxicological Limits on Airborne Concentrations of Transportable (soluble) Uranium Agency Chronic Exposure Occupational Limit, mg/m3 Reference NRC 0.2 Footnote to Appendix B, 10 CFR 20 (NRC 1992a) ACGIH 0.2 Threshold Limit Values and Biological Exposure Indices for 1997, American Conference of Governmental Industrial Hygienists (ACGIH 1997) OSHA (a) 0.05 (soluble) 0.25 (insoluble) 29 CFR 1910.1000 NIOSH 0.05 National Institute for Occupational Safety and Health (a) Preferred/recommended limit. Past limits for single acute inhalation intakes have been set by the International Commission on Radiological Protection in its Publication 6, (ICRP 1964) to 2.5 mg of soluble uranium inhaled in any one day. This value is based on one day’s intake at the maximum permissible concentration (at the time) of 210 µg/m3. Lawrence (Lawrence 1984) derived acute inhalation intake limits of 15 and 80 mg for Class D and Class W materials, respectively. This derivation is based on not exceeding a kidney burden of 3 µg U/g kidney after a single acute inhalation. NRC regulations at 10 CFR 20 limit the intake of soluble uranium to 10 mg in a week. Chronic exposure to a concentration of 0.2 mg/m3 results in a weekly intake of 9.6 mg (40 h/week x 1.2 m3/h x 0.2 mg/m3) and a steady-state kidney burden of roughly 900 µg, when the ICRP Publication 30 metabolic model for Class D uranium is used. This same model indicates that an acute intake of 18 mg will result in a prompt kidney burden of approximately 900 µg. However, 10 CFR 20 limits acute exposures to 40 DAC-hours, or 9.6 mg.

Recurrent concerns have arisen about the adequacy of existing limits intended to prevent chemical damage to kidneys. These concerns have focused particularly on the • lack of data on the effects of combined exposures to UO2F2 and HF • lack of detailed information on effects of short-term exposures to soluble/transportable uranium in the range from 100-1000 mg/m3 • lack of data on thresholds for repairable injury. DOE sponsored research to determine the exposure levels that would be expected to 1) have no effect, 2) cause non-lethal injury, and 3) be lethal to 50% of the exposed population (LD 50). Researcher consensus resulted in the uptake levels (in µg U/g) listed in Table 2-9 along with the corresponding total uranium in 70-kg standard man. 2-22 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Table 2-9. Uranium Levels for Various Effects Effect Uranium Absorbed into Bloodstream, µg U per g body weight Corresponding Class D Uranium Intake in Standard Man, mg Corresponding Kidney Burden µg U/g Kidney No effect 0.04 5.9 1.1 Maximal Nonlethal 0.08 11.6 2.2 LD50 2.0 294 54.8 The values for "standard man" are based on the ICRP Publication 30 model for uranium metabolism (47.6% of inhaled Class D uranium is taken up into the bloodstream, and 12% of that goes to the kidneys). For example, the "no effect" value in Table 2.9 corresponds to a kidney burden of (5.9)(.476)(.12) = 0.337 mg. The mass of kidney tissue in standard man is 310 g, so this kidney burden represents 1.1 µg uranium per gram of kidney tissue. An airborne contamination limit from this "no effect" kidney burden can be derived by calculating the airborne uranium concentration at which chronic exposure would result in a kidney burden that just equals the "no effect" burden. In the illustrative analyses below, the 1500-day component of ICRP Publication 30's kidney retention function is neglected, since this contribution is neglig ible. For chronic exposure to a constant concentration, the maximum kidney burden will occur at the equilibrium condition--when the amount of uranium entering the kidney each day equals the amount being removed from the kidney. The daily kidney uptake rate and removal rate are calculated from the following formulas: K = Br x Ca x fb x fk where K = kidney uptake rate (mg/day) Br = breathing rate (m3/day) Ca = air concentration (mg/m3) fb = inhaled fraction entering bloodstream (0.476) fk = bloodstream fraction entering kidneys (0.12) and R = λKb where R = kidney removal rate (mg/day) λ= 0.693/T1/2 (day-1) Kb = amount in the kidney (mg) T1/2= biological half-life of U in kidney = 6 days To calculate the concentration at which chronic exposure would result in a kidney burden of 0.337 mg, the uptake rate in kidney is set equal to the removal rate for a 0.337-mg kidney burden: R = (0.337) x 0.693/6 = 0.039 mg/day K = Br (m3/day) x Ca(mg/m3) x (0.476) x (0.12) K= R = 0.039 mg/day 2- 23 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities BrCa x (0.476) x (0.12) = 0.039 mg/day Br x Ca = 0.68 mg/day Standard man breathes 9.6 m3 of air in an 8-hour day, so the resulting concentration limit is 0.68/9.6 = 0.07 mg/m3. This is 40% higher than the OSHA standard for soluble uranium of 0.050 mg/m3. Consequently, the OSHA limit is somewhat conservative for exposures to soluble/transportable (i.e., Class D) uranium.

2.4.1 Human Response Indicators Most data on human response to uranium exposure comes from accidental exposures (generally UF6 releases). Accidental exposures to UF6 have resulted in fatalities on at least three occasions. The primary cause of injuries and fatalities has been HF that was formed by hydrolysis of UF6, rather than exposure to UF6 itself. Several individuals who received high, non-fatal exposures experienced pulmonary edema, nausea, vomiting, abdominal cramps, and chemical burns on the skin due to HF exposure. In addition, urinary abnormalities, such as transient albuminuria (albumin in urine) and the presence of red cells and casts, were observed, as was retention of nitrogenous products such as urea and non-protein nitrogen in the blood. The urinary and blood abnormalities are indicators of kidney damage, and are the result of inhibited resorption in the tubules. Animal studies indicate that urinary abnormalities can be observed after exposures that are well below lethal levels. In addition, urinary abnormalities such as proteinuria (protein in urine), glucosuria (glucose in urine), and polyuria (increased urine volume) have all been observed following uranium exposure, as has the presence of certain enzymes in urine. Of all these abnormalities, glucosuria appears to be the most sensitive and most nearly proportional to uranium exposure. Once absorbed into the blood, uranium is distributed to bone and kidneys, with a portion of the uptake being generally distributed throughout the body. For inhaled uranium, residence time in the lungs depends upon the solubility of the material. Material that is deposited in the lungs is cleared via the bloodstream, the pulmonary lymph, and the gastrointestinal (GI) tract. Approximately 1 % of the uranium is absorbed into the bloodstream from the GI tract. In the event of an acute exposure to highly transportable (Class D) uranium compounds, urine samples should be collected 3-4 hours post-exposure and analyzed for uranium as soon as possible. If the uranium concentration is less than 2.0 mg/L, it is unlikely that any significant kidney damage has occurred or will occur. However, it is important to check the urine for biological indicators of damage at any exposure above 2.0 mg/L. While the most sensitive indicators are increased volume and glucose levels, these are useful only if data on what is "normal" for the individual involved are available. Lacking that information, it is best to check for albuminuria as an indicator of kidney damage. If kidney damage is suspected, a specialist in urinary disorders should be consulted. In general, a urine uranium level greater than 6.0 mg/L will produce some level of albuminuria. A level of 20 mg/L indicates a very serious exposure with potentially life-threatening consequences and would indicate the need for immediate hospitalization. 2-24 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 2.4.2 Transfer to the Fetus Little information exists on the placental transfer or developmental toxicity of uranium isotopes (Sikov et al 1992). The data available with pregnant rats suggest that the effects produced from exposure to uranium may be due to chemical toxicity to the pregnant animals and their embryos/fetuses. Fetoplacental concentrations of uranium peak one day following intravenous injection of a pregnant rat. Although concentrations in the placenta decrease thereafter, the concentration in the fetal membranes remains relatively constant. Selective deposition in some fetal organs will occur when exposure is during the fetal developmental stages (NRC 1992b).

Data from animal experiments suggest that the distribution pattern of uranium is fairly uniform, especially at the early stage of gestation. Concentrations of uranium in the embryo/fetus are taken to be the same as those in the maternal soft tissues (excluding the kidney) during the first two months, and they progressively increase thereafter. Following transfer into the embryo-fetus, uranium activity is assumed to be distributed uniformly and to remain without excretion. 2.5 CHEMICAL VERSUS RADIOLOGICAL HAZARDS Both the chemical and radiological hazards of uranium are moderate compared to those of other industrial materials and radionuclides. Table 2-6 provides 10 CFR 835 derived air concentration values for selected radionuclides. Table 2-10 compares Threshold Limit Values (TLV) published by ACGIH for uranium and selected other metals. The comparison of TLVs is presented to provide perspective on the need for uranium workplace controls, as compared to other hazardous materials. Since these materials affect the body in different ways, this should not be considered a comparison of relative hazards. The predominant hazard associated with uranium exposure depends upon its degree of enrichment, its chemical form, and its physical form. The degree of enrichment determines the gamma radiation intensity and the overall specific activity. The effect that enrichment has on specific activity is illustrated in Figure 2-2. That figure (adapted from NRC Regulatory Guide 8.11) also gives 3.6 x 10-7 Ci/g as the specific activity of depleted uranium and lists the formula used in Section 2.1.1 for calculating specific activity of enriched uranium. Table 2-10. 1999 ACGIH Threshold Limit Values (TLVs) for Selected Metals Soluble and Insoluble TLV Metal TLV-TWA, mg/m3 TLV-STEL, mg/m3 Uranium 0.2 0.6 Beryllium 0.002 -- Lead 0.05 0.45 Mercury vapor, all forms except alkyl 0.05 -- Arsenic 0.01 -- TLV-TWA = Threshold Limit Value, Time-Weighted Average TLV-STEL = Threshold Limit Value, Short-term Exposure Limit 2-25 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities The relative activities of the primary uranium isotopes are also significantly affected by the degree of enrichment (see Figure 2-2). The figure shows that total activity is due chiefly to 238U for depleted and 234U for enriched uranium, while 235U accounts for little of the total activity, even at very high enrichments. Chemical form determines solubility and consequent transportability in body fluids. ICRP Publication 30 classifies all materials into three inhalation classes--D, W, and Y (soon to be Types F for fast, M for moderate, and S for slow). Class D is most transportable (pulmonary removal half-time of days), Class Y the least transportable (removal half-time of years), and Class W an intermediate category (removal half-time of weeks). The transportability of an inhaled or ingested material determines its fate within the body and, therefore, the resulting radiation dose or chemical effect. Table 2-11 lists several common uranium compounds and their assigned transportability classes. Table 2-11. Inhalation Classification for Some Uranium Compounds Uranium hexafluoride UF6 Class "D"(a) Uranyl fluoride UO2F2 Class "D"(a) Uranyl nitrate UO2(NO3)2 Class "D" Uranyl acetate UO2(C2H3O2)2 Class "D" Uranyl chloride UO2Cl2 Class "D" Uranyl sulfate UO2SO4 Class "D" Uranium trioxide UO3 Class "W" Uranium tetrafluoride UF4 Class "W"(a)

Uranium oxide U3O8 Class "Y"(b) Uranium dioxide UO2 Class "Y"(b) Uranium tetroxide UO4 Class "W" Ammonium diuranate (NH4)2 + U2O7 Class "W"(b) Uranium aluminide UAlx Class "Y"(a) Uranium carbide UC2 Class "Y" Uranium-zirconium alloy UZr Class "Y" High-fired uranium dioxide UO2 Class "Y"(b) (a) "D" and "W" and "Y" are inhalation solubility classes established by the ICRP: "D" class material is very soluble, with lung retention time in days; "W" class material is moderately soluble, with lung retention time in weeks; "Y" class material is relatively insoluble, with lung retention time in years. (b) Ammonium diuranate is known to contain uranium as UO3, and should not be assigned to a single inhalation class. The solubility of uranium oxides is very dependent on heat treatment. The rate of oxidation may also affect the solubility. Although references assign inhalation classes to various uranium compounds, it is recommended that solubility studies be performed to characterize the actual materials present. 2-26 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities This listing is intended to provide general guidance only, as a given material’s transportability will depend upon a number of parameters including its processing history. It is recommended that each facility determine the transportability of materials it handles using one of the accepted techniques. Physical form influences potential hazards since non-dispersible forms generally do not constitute an ingestion or inhalation hazard. Because inhalation of uranium potentially poses both radiological and toxic hazards, one must determine which hazard is most limiting and whether or not either hazard can be ignored under certain circumstances. When radiological hazards are limiting, chemical hazards can generally be neglected, except in overexposure situations. When chemical hazards are limiting, radiological hazards can be neglected only if radiation doses are below regulatory concern. Radiological monitoring is required by DOE for individuals who are likely to exceed 100 millirem CEDE in a year. Therefore, it is prudent to calculate organ doses and CEDE for all confirmed intakes, since additional exposures in the same year may result in a total dose exceeding the mandatory individual monitoring threshold. Even in low-potential exposure level situations, a comprehensive dosimetry/control program can prove invaluable in public relations concerning possible future legal litigation. The limiting hazard (chemical or radiological) depends on the transportability (solubility in body fluids), enrichment, and duration of exposure (acute or chronic). As discussed in Section 2.4, the "no effect" value of intake corresponds to a kidney burden of 0.337 mg. The 0.337 mg kidney burden and ICRP Publication 30 metabolic models are used in the following examples to determine the relative hazards for acute exposure situations. The 0.337 mg kidney burden corresponds to a chronic exposure of 0.07 mg/m3. OSHA exposure limits for uranium are 0.05 mg/m3 for soluble forms and 0.25 mg/m3 for insoluble forms. These exposure limits are used to determine the relative hazards for chronic exposure situations. For radiological considerations, soluble forms of uranium are considered to be Class D and insoluble forms, Classes W and Y.

To determine which hazard is limiting for an acute exposure, the intake corresponding to "no effect" kidney burden is first calculated and appropriate annual limit on intake (ALI) determined. The formula for specific activity is solved in order to determine the enrichment at which the "no effect" intake is equal to one ALI. For chronic exposure scenarios, the OSHA exposure limit and appropriate derived air concentration (DAC) are used. The formula for specific activity is solved to determine the enrichment at which the DAC is equal to the OSHA limit. These enrichments form the "dividing line" between chemical and radiological effects as the limiting hazard. Exposures to higher enrichments are limited by radiological effects; exposures to lower enrichments by chemical effects. Example 3a provides the methodology for determining the "dividing line" enrichment for the acute exposure scenario. Example 3b provides the methodology used for the chronic exposure scenario. The following variables are used in these examples: fb = fraction of inhaled uranium that promptly enters the bloodstream fk = fraction of uranium in bloodstream that enters kidneys SA = specific activity of uranium in microCi/g obtained from ALI/intake or DAC/concentration Br = breathing rate for standard man = 2,400 m 3/year 2-27 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Table 2-12 shows the values used for fb and fk, ALI, and the resulting "dividing line" enrichments for acute and chronic exposures. Several aspects of these derivations must be kept in mind when using this information. First, the derivation is based on standard metabolic models and therefore does not necessarily reflect the effects of a uranium uptake on a real person. Because individual metabolism’s will not necessarily agree with the model, the enrichment at which chemical and radiological effects are equally limiting cannot be precisely determined. Uncertainty in the relationship between enrichment and specific activity introduces additional imprecision. Consequently, exposures for both chemical and radiological impact for uranium uptakes at enrichments near the calculated "dividing line" enrichment should be evaluated. Table 2-12. Determination of "Dividing Line" Enrichments Above Which Radiological Monitoring Requirements Become Limiting Class fb fk Annual Limit on Intake µCi Specific Activity of "Dividing Line" Enrichment for Radiological Dose Limit "Dividing Line" Enrichment for Radiological Dose Limit "Dividing Line" Enrichment for 2% Monitoring Threshold D 0.476 0.12 1 169.5 µCi/g (b) 7.38% W 0.12 0.12 0.7(a) 29.9 µCi/g 52.8% 0.52% Y 0.05 0.12 0.04 0.71 µCi/g 0.82% (c) (a) ICRP Publication 30 lists Class W ALIs of 0.7 µCi for 234U and 0.8 µCi for 235U and 238U. This difference is the result of rounding to one significant figure. Non-rounded values for the three isotopes are all approximately 0.75 microCi. (b) The resulting enrichment is greater than 100 %. Consequently, chemical toxicity is limiting for acute exposures to Class D uranium approaching the radiological dose limit. (c) The resulting enrichment is lower than that of depleted uranium. Consequently, radiological concerns are limiting for acute exposures to Class Y uranium at the monitoring threshold. The impact of the requirement to perform individual radiological monitoring at 2% of the regulatory dose limits can be assessed by reducing ALIs by a factor of 50, then repeating the calculations described in Examples 3a and 3b. Table 2-13 summarizes the results of these calculations.

The effects that enrichment, chemical form, and physical form have on the hazards associated with uranium are summarized in Table 2-13. The comparison of relative chemical and radiological hazards is based on a derived kidney burden resulting from an acute exposure at the "no effect" threshold. The effect of using the OSHA exposure limits of 0.05 mg/m3 for soluble forms of uranium (Class D) and 0.25 mg/m3 for insoluble forms (Classes W and Y) is shown for chronic exposures. The derivations used here can be applied to any limit on radiological or chemical toxicity, be it a regulatory or an internal dose control limit. It should be emphasized, however, that the radiological impact should be considered for all intakes, even for exposure situations where chemical toxicity is limiting. 2-28 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Table 2-13. Impact of Requirement To Monitor at 2% Enrichments above which radiological concerns predominate Acute Chronic Transportability Class Using 100% of Radiological Limit Using 2% of Radiological Limit Using 100% of Radiological Limit Using 2% of Radiological Limit D (1) 7.38% 18% (2) W 52.8% 0.52% 12.8% (2) Y 0.82% (2) (2) (2) (1) Chemical toxicity concerns are limiting at all enrichments. (2) Radiological effects are limiting at all enrichments. Example 3a - General Solution, Acute Exposure Step 1. Determine the intake that results in a kidney burden of 0.337 mg: Step 2. Use the quadratic formula and equation for determining specific activity to calculate the enrichment that corresponds to the specific activity obtained in Step 1. Step 3. One solution will be less than zero. The other will be the enrichment that is the “dividing line” between chemical and radiological effects. 2-29 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Example 3b - General Solution, Chronic Exposure Step 1. Determine specific activity at which chronic exposure results in being exposed to one Derived Air Concentration (e.g., ALI divided by breathing rate) at the OSHA exposure limit. The ALI for Class D is used with the OSHA exposure limit for soluble forms of uranium. The ALIs for Classes W and Y are used for the OSHA exposure limit for insoluble forms of uranium. Step 2. Use quadratic formula and equation for determining specific activity to calculate enrichment which corresponds to the specific activity obtained in Step 1. Step 3. One solution will be less than zero. The other will be the enrichment that is the “dividing line” between chemical and radiological effects. If both solutions are less than zero, then radiological effects are always limiting. 2-30 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 2.6 INDUSTRIAL HAZARDS The principal industrial hazards associated with uranium are fires, hydrogen generation, generation of oxides of nitrogen, and associated mechanical hazards characteristic of heavy objects, i.e., back injuries from lifting, dropping heavy parts on feet, etc. Hydrogen fluoride (HF) and oxides of nitrogen (NOx) are by-products or reactants of common chemical processes. Hydrogen (H2) can be generated by reaction of water with uranium metal, and finely divided uranium or uranium chips with a large surface area to volume ratio can ignite spontaneously. 2.6.1 Hydrogen Fluoride

Hydrogen fluoride is an extremely corrosive acid that is relatively volatile in its anhydrous form. Anhydrous HF is a reactant for the production of UF4 from UO3, a by-product of the production of UF4 from UF6, and is generated whenever UF6 is released to the atmosphere (H20 in air + UF6 → UO2F2 and HF). External contact with HF results in chemical burns of the skin, while exposure to airborne HF causes chemical burns/irritation of the eyes, nose, and throat. Significant inhalation can result in pulmonary edema. Chronic exposure to excessive fluoride concentrations results in increased radiographic bone density and may eventually cause fluorosis (osteosclerosis). In general, individuals can smell HF at levels of 0.02-0.2 mg/m3, much lower than the TLV of 2.5 mg/m3. The TLV was set based primarily on the irritation of eyes and mucous passages rather than on permanent damage. Because an airborne concentration of 10 mg/m3 is intolerable, personnel exposed to such levels will evacuate the area if they are able to do so. Exposure for as little as 15 minutes to an airborne concentration of 20-30 mg/m3 may prove fatal (pulmonary edema). The AIHA Emergency Response Planning Guides (ERPGs) for HF are as follows: ERPG-3, 42 mg/m3; ERPG-2, 17 mg/m3; and ERPG-1, 4 mg/m3. The NIOSH IDLH value is 25 mg/m3. 2.6.2 Nitric Compounds Nitric acid is widely used for digesting uranium metal and uranium-bearing compounds and for "pickling" metal products to inhibit oxidation. Concentrated nitric acid gives off fumes that cause irritation to eyes, mucous membranes, and skin. Significant inhalation can result in pulmonary edema. The ACGIH TLV-TWA and TLV-STEL values for nitric acid are 2 ppm and 4 ppm, respectively. When uranium materials, especially metal, are dissolved in nitric acid, oxides of nitrogen (NO x) are generated. The term NOx is applied to mixtures of nitric oxide (NO) and nitrogen dioxide (NO2). The ACGIH TLV-TWA and STEL are 25 ppm and 35 ppm, respectively. Exposure to NO2 can cause eye irritation, coughing, mucoid frothy sputum, shortness of breath, chest pain, pulmonary edema, cyanosis, tachypnea (abnormal rapid breathing), and tachycardia (abnormal rapid heartbeat). 2.6.3 Hydrogen Gas Hydrogen gas (H2) is used as a reactant in the production of UF4 from UF6 and in the reduction of UO3 to UO2, an intermediate step in the production of UF4 from UO3. The H2 is usually generated by dissociating ammonia, so associated ammonia rather than hydrogen is frequently identified as the reactant in those processes. Any facility where H2 is used as a reactant should include design features (e.g., H2 monitors, roof vents, etc.) to ensure that hydrogen accumulations do not occur. Generally, H2 hazards and control features are identified in facility Documented Safety Analyses. Hydrogen can also be generated when moisture contacts uranium metal, especially finely divided uranium metal such as 2-31 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities machining chips. Care must be taken to ensure that H2 generated in this manner does not accumulate (in closed drums or storage containers for example). 2.6.4 Fire

Finely divided uranium metal is highly reactive or pyrophoric, capable of igniting spontaneously. This type of material should be handled and stored in a manner that minimizes fire potential. Typically, machining chips are stored under water or machining oil in open storage containers so that any H2 generated does not accumulate. Neither water spray, CO2, nor halon extinguishers are effective in fighting uranium fires. In fact, halon may be explosive if directed at burning uranium and can produce very toxic fumes and gases. Small uranium fires can be smothered in MET-L-X powder (a mixture of sodium chloride and potassium carbonate). Larger fires, involving drums of machining turnings, for example, can be controlled by immersing the burning container in water. Even this will not immediately extinguish the fire because the hot uranium metal dissociates the water into H2 and O2, providing fuel and oxygen for the fire. If the quantity of water is sufficient, eventually the water will provide enough cooling to extinguish the fire, but a significant amount of water can boil away in the process. If the water level is allowed to fall low enough to uncover the uranium while the fire is still burning, it will resume burning visibly. DOE-HDBK-1081-94, Primer on Spontaneous Heating and Pyrophoricity, (DOE 1994e) contains additional guidance. 2-32 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.0 RADIATION PROTECTION An effective radiation protection program at a uranium facility requires scrupulous attention to controlling both internal and external doses. The radiation protection program should ensure the detection and quantification of all types of radiation (i.e., alpha, beta, neutron, gamma, and x-ray) over wide energy ranges. The radiation detection instruments should be properly calibrated and routinely checked. Emphasis should be on establishing controls for internal and external radiation exposure using ALARA guidelines. Prompt and accurate assessment is important in determining each individual’s dose and in establishing an accurate historical record. This section defines the basis for establishing a comprehensive radiation protection program. 3.1 REGULATIONS AND STANDARDS DOE has established occupational radiation protection regulations in 10 CFR 835, Occupational Radiation Protection. DOE has provided supporting and clarifying guidance in the DOE G 441.1 series of Guides, DOE-STD-1098-99, Radiological Control, and DOE Radiological Control Technical Positions. Other related source documents include publications of the EPA, ANSI, ICRP, NCRP, and UNSCEAR. Individual states may also have their own radiological control regulations, with equivalent or more restrictive requirements than the Federal regulations. 3.2 RADIATION PROTECTION PROGRAMS An effective radiation protection program consists of a group of related and integrated functional elements. The documentation that describes the DOE activity’s program to control occupational radiation protection is referred to as the documented radiation protection program (RPP). Although the actual titles and contents of the functional elements are left to the discretion of DOE’s operating entities, DOE G 441.1-1A, Management and Administration of Radiation Protection Programs (DOE 2003a), suggests the following, based on the content of 10 CFR 835:

• Organization and Administration • ALARA Program • External Dosimetry Program • Internal Dosimetry Program • Area Monitoring and Control • Radiological Controls • Emergency Exposure Situations • Nuclear Accident Dosimetry • Records • Reports to Individuals • Radiation Safety Training Each of these functional elements is discussed in more detail below. 3.2.1 Organization and Administration This functional element addresses the overall administration of the program, including the documented RPP itself, various organizational and institutional issues, and program assessment. DOE G 441.1-1A and the RCS provide detailed guidance on implementing these requirements. 3-1 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Although 10 CFR 835.101 requires that DOE activities be conducted in compliance with a documented RPP, the rule does not establish specific requirements for RPP format and content. Due to the wide range of activities undertaken by and for DOE, there is significant flexibility in these provisions. Cognizant DOE line management determines the acceptable format and content of the documented RPP. However, the documented RPP shall address each requirement of 10 CFR 835 and shall be approved by DOE (10 CFR835.101). Any changes that decrease the effectiveness of the RPP shall be approved by DOE before implementation (10 CFR835.101). Internal audits of the RPP, including examination of program content and implementation, shall be conducted through a process that ensures all functional elements are reviewed no less frequently that every 36 months (10 CFR 835.102). An effective quality assurance program for radiation protection should include establishment of appropriate standards of performance for essential activities and equipment, with an effective system of documentation and traceability of those activit ies and of the use of the equipment. Proper maintenance of those records will be necessary for reference purposes. Additional requirements and guidance are provided in 10 CFR 830, Nuclear Safety Requirements (DOE 2001d), DOE O 414.1B (DOE 2004), Quality Assurance, and their associated guides. Specific guidance applicable to RPPs is provided in DOE G 441.1-1A. 3.2.1.1 Administrative Controls In any facility that handles radioactive materials, the major controls protecting workers, the public, and the environment are physical design features, such as structures and installed equipment, that shield, contain, and confine the radioactive materials. However, to allow useful work to be performed in the facility and to ensure that its protective features remain effective, a number of administrative controls are ordinarily required. These controls are usually described in and implemented through a series of policy statements and procedures related to the operations and maintenance activities to be carried out in the facility. All personnel who work in controlled areas should be familiar with the administrative controls that apply to their work. Changes or additions to administrative controls should be effectively communicated to all persons who may be affected. Radiation Protection Procedures A uranium facility should have a written policy on radiation protection, including a policy on keeping exposures ALARA.

To ensure facility activities are executed safely and in a manner that consistently meets management expectations, documented procedures should provide detailed instructions for implementing various functional elements of the RPP. Written procedures shall be developed and implemented as necessary to ensure compliance with 10 CFR 835, commensurate with the radiological hazards created by the activity and consistent with the education, training, and skills of the individuals exposed to those hazards (10 CFR 835.104). Responsibilities and actions required of management and workers should be clearly and unambiguously stated. It is not necessary for written procedures to be developed and implemented for all of the requirements of 10 CFR 835. Written procedures should be developed and employed under the following circumstances: − Worker health and safety are directly affected; − the expected outcome for the process or operation requires that a specific method be followed; 3-2 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities − the process or operation is infrequently used and competence training cannot assure adequate implementation; or − to document the approved method to implement specific processes or operations. In evaluating the need for written procedures, consideration should be given to the level and extent of the radiological hazards, the complexity of the measures required to achieve compliance, and the education, training and skills of the individuals who must implement those measures. Under such a regimen, a low hazard activity employing a stable staff of highly educated and skilled workers having demonstrated an advanced knowledge of radiation protection principles and practices could have fewer and less detailed procedures than a higher hazard activity employing a transient workforce with less knowledge of radiation protection practices and principles. The DOE G 441.1 series of Guides provide additional guidance regarding specific procedural aspects of the RPP. All radiation protection procedures and controls should have formal, recognizable technical bases for limits, methods, and personnel protection standards. Procedures should be adequately documented, updated periodically, and maintained in a centralized historical file. A control system should be established to account for all copies and ensure all new procedures are included in the historical files. A designated period of time for maintaining historical files should be established. ANSI/HPS N13.6, Practice for Occupational Radiation Exposure Records Systems (ANSI/HPS 1999) provides guidance on maintaining historical files. In addition, radiation protection procedures should have a documented approval system and established intervals for review and/or revision. A tracking system should be developed to ensure that the required reviews and revisions occur. Guidance for writing procedures can be found in DOE/NE/SP-0001T, Writer’s Guide for Technical Procedures (DOE 1991a). Management Commitment Management commitment to safety is the most important characteristic of an effective radiological control program. If the management commitment to safety is strong, the radiological control program will be valued and respected. The radiological control program should be provided adequate authority to permit performance of necessary assignments and program implementation. Management commitment to the ALARA concept is particularly important (see Article 111 of the RCS). Adequate personnel, equipment, and funding should be available as a part of this commitment.

Radiological Control Organization The radiological control organization should be structured so that all of the activities required to provide support to line management and workers can be accomplished. Radiological Control Organization Independence and Reporting Level The radiological control organization should be independent of the line organization responsible for production, operation, or research activities and should have an equivalent reporting level. Because radiological control personnel should have the authority to balance operations with safety, they should not report directly to the administrators of operations. When shift work is involved, the operations shift supervisor may make minor radiological control decisions in support of the shift’s Radiological Control Technicians (RCTs); however, decisions involving basic policies and procedures should be directed to a separate radiological control organization. 3-3 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities If a safety organization includes the radiological control program, it must be high enough in the company to allow direct access to the company president or equivalent. If the radiological control program is administered by a separate radiological control organization, that organization must also be in a position to have direct access to the company president. This is to safeguard the program from the pressures of production that exist in the operational environment and to keep it independent of operating organizations. A system of guides, policies, and procedures should be established to clearly identify the inter- relationships, responsibilities, and authorities of those involved with the development, operation, and maintenance of the facility and the health and safety of the employees. These guides, policies, and procedures should be documented and reviewed at least once every year. Adequacy of Personnel and Equipment A sufficient number of qualified and, where required, certified radiological control personnel must be available to perform necessary tasks for support of uranium facility startup and operation. Sufficient equipment, including protective clothing, respiratory protective equipment, and radiation detection instrumentation should be available to support RCTs and operating personne l in the perfor- mance of work in controlled areas. Staffing and Staff Qualifications A cadre of operating and maintenance personnel who have experience in the operation of a uranium facility should be established during the construction of a new facility. The remainder of the operating and maintenance staff should be hired as soon as possible and should receive formal and informal training from the experienced personnel. This step is extremely important to enable all personnel to grow with the facility and learn the details of the operations. Once operations start, potential problems already should have been identified, and engineering or administrative changes should have been made to resolve them. Staffing in the radiological control organization requires technicians and professionals in many support areas. A successful radiological control program is highly dependent upon the availability of adequate staff support in disciplines such as environmental monitoring, instrument maintenance and calibration, internal and external dosimetry, meteorology, safety analysis, and risk management. Radiological Control Technician Training

A thorough RCT training program should be established at uranium facilities. Before uranium operations begin, a trained and qualif ied staff of RCTs should be present. All RCT training should be accomplished in accordance with the RCS and DOE-HDBK-1122-99, Radiological Control Technician Training Program (DOE 1999d). Professional Staffing and Qualifications The senior staff of the radiological control organization should include health physicists and other professionals with four-year degrees in science or engineering. A continuing training program should be established for facility personnel. Pursuit of certification by the American Board of Health Physics for senior and professional staff members should be encouraged. At least one professional staff 3-4 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities member at the uranium facility should have a minimum of three years of radiological control experience in the operation of uranium facilities. Technician Staffing and Qualifications Recommendations for minimum entry-level requirements for RCTs are given in the RCS and the Radiological Control Technician Training Program. They include a high school education or equivalency and knowledge of certain scientific fundamentals. If a two-year degree in nuclear technology or an equivalent discipline is locally available, completion of such a program should be encouraged. Where possible, the RCTs and other members of the radiological control staff should have a minimum of one year’s experience working at a uranium facility. Such experience is an important prerequisite to allow them to work unsupervised. Personnel hired without such experience should work an internship of six months under the leadership of a qualified RCT or supervisor with experience in that facility. RCTs should be encouraged to pursue registration by the National Registry of Radiation Protection Technologists. Training Staff Qualifications All training instructors and materials should meet the requirements of DOE Order 5480.20A Ch.1. The RCS provides additional guidance. Each uranium facility should develop performance-based training that reflects radiological conditions present at the facility. This training should be monitored to ensure that site-specific, worker-performance-based measures, and practical factors are included in the uranium training. Health Physicist Training Involvement Facility health physicists should have comprehensive knowledge of all of the material on uranium radiation safety included in the training programs for radiation workers and RCTs. In addition to the previously discussed RCT training material, DOE has developed several other radiation safety training courses and qualification standards which may provide useful information. These documents include: DOE-HDBK-1113-98, Radiological Safety Training for Uranium Facilities (DOE 1998c) DOE-HDBK-1130-98, Radiological Worker Training (DOE 1998d) DOE-HDBK-1131-98, General Employee Radiological Training (DOE 1998e) DOE-STD-1107-97, Knowledge Skills and Abilities for Key Radiation Protection Positions at DOE Facilities (DOE 1997) Staffing Levels At least one professional health physicist is recommended to be on the staff of each major uranium facility as a full-time employee. There is no rule of thumb for determining the number of RCTs needed for a given uranium facility. 3-5 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities

The number of RCTs should be based on an analysis that provides for sufficient coverage on each shift, given the number of samples, surveys, and other work to be performed, the time of training, donning and doffing of protective clothing, shift turnover procedures, and other similar considerations. The dose rate and individual dose limits in the facility may also lead to the need for additional personnel. Consideration should be given to having sufficient numbers of personnel to respond to off-normal conditions and emergencies as well as routine work. Major maintenance, modifications, or decommissioning activities may require additional personnel. 3.2.2 ALARA Program The policy for maintaining radiation exposures ALARA has existed in principle since the early 1940s. The evolution of ALARA into a formal program began in the early 1960s. Although there is, and has been since the 1940s, a series of official established dose limits, they do not represent ALARA. ALARA is a continuous process of controlling and managing radiation exposure to workers, the general public, and the environment. Although ALARA is based upon protection of people and the environment, the philosophy is also grounded on sound economic and operating principles. The responsibility for maintaining radiation exposures ALARA is not a unique responsibility of management or radiological control personnel. It is a responsibility of everyone involved in managing, supervising, or performing radiation work. It is imperative to teach administrative personnel to support the principles and practice of ALARA, and to train all workers to consider ALARA as they prepare for and perform their work. 3.2.2.1 Assignment of ALARA Responsibility and Authority Limiting radiation exposures to the lowest levels commensurate with economics and the work to be accomplished has long been a part of radiological control and radiological protection programs of DOE and its contractors. 10 CFR 835 establishes the policy of maintaining ALARA doses for workers and the public resulting from radiation from DOE operations. Plans and programs are required to be prepared and implemented, and records must be maintained to demonstrate the implementation of ALARA. DOE G 441.1-2, Occupational ALARA Program Guide (DOE 1999e), the RCS, and PNL-6577, Health Physics Manual of Good Practices for Reducing Radiation Exposure to Levels That are as Low as Reasonably Achievable (PNL 1988a), provide additional guidance. An ALARA committee should be established at the uranium facility. The membership should include managers and workers from the line, the technical support organization, and the radiological control organization. A line manager, such as a Director of Operations, Research, or Maintenance, should serve as the committee chair. The ALARA committee should make recommendations to management to improve progress toward minimizing radiation exposure and radiological releases. 3.2.2.2 Current Status of ALARA Programs Currently, it is common practice in DOE facilities to have a well-structured ALARA plan for the entire facility, with more detailed plans in the various buildings or functional subunits of the facility. There is ordinarily a facility coordinator who administers the overall ALARA plan and reports to top-level management of the facility. Coordinators for the various buildings or subunits of the facility receive guidance from the overall facility coordinator and report the results of their ALARA programs to that individual.

3-6 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.2.2.3 Achievement of Goals To ensure improving radiological performance, at the beginning of each year, each facility should prepare radiological performance goals. At intervals commensurate with the radiological risk, the contractor should provide DOE with an interim status report of the goals. At the end of the calendar year, the contractor should provide DOE an Annual Goal Status Report. Identifying specific ALARA goals in uranium facilities requires close coordination between the facility ALARA team members (operations, maintenance, and radiological control personnel) made up from a cross-section of personnel representing the various work elements of the facility. ALARA goals may be formulated as qualitative or quantitative types of goals, but must be measurable and achievable, with clearly defined endpoints. 3.2.2.4 Quality Assurance Important aspects of any ALARA program are the measurement of beneficial effects and the determination that important factors, such as economic impacts, the time involved in accomplishing tasks, and the utilization of personnel, are being optimized. To accomplish these objectives, it is necessary to have a written plan for the ALARA program and high quality records of activities involving exposures to workers, the public, and the environment. These permit comparisons with past experiences and analysis of the recorded activities. In many cases, such studies of the recorded activities not only confirm satisfactory execution of the work, but reveal opportunities for future improvements. One approach that works well is the inclusion of an ALARA worksheet with the RWP. Such a worksheet should be prepared by an individual with responsibilities for the work to be performed, a relatively detailed knowledge of the radiological conditions, and knowledge of what is required to accomplish the task. The worksheet should contain estimates of the time to complete the task and the expected radiation doses to be received. If any specially-engineered devices are used to control personnel exposure, they should be noted on the ALARA worksheet, with any special instructions they require. These worksheets provide valuable information for analysis of the effectiveness of the ALARA program for each job. 3.2.2.5 Technical Aspects The technical aspects of ALARA programs include not only the standard equipment regularly used in controlling dose to workers, the public, and the environment, such as facility shielding, ventilation filters, installed and portable radiation measuring instruments, but also many special devices that may be used temporarily. Special devices can be used to provide exposure control and/or containment when it may not be practical without them. These include temporary shields, tents or greenhouses, portable fans, ductwork and filters, and special fixtures to hold highly radioactive materials requiring detailed inspection, repair, modification, or fabrication. Such devices can permit doing difficult work at low radiation doses, which might not be possible otherwise. Some of these special devices may have general application and be kept on hand for use as needed. In some cases, devices would have to be specially fabricated for a specific task. Because this would ordinarily have a significant effect on the cost of doing that job, the economic aspects of doing or not doing the job should be carefully evaluated.

3-7 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.2.2.6 Attributes of Effective Review and Audit Evaluation of the effectiveness of an ALARA program requires both reviews and auditing. The reviews will include detailed examination of the written ALARA program plan and the records of ALARA activities. The objectives in such reviews are to find if the written plan is being followed, and what is working or not working well. Such reviews can be performed adequately by either a knowledgeable member of the facility staff or an equally knowledgeable outsider. The written report of a review should be directed to a member of management who is responsible for implementation of the ALARA program. Audits are best performed by an outside expert who is knowledgeable about work with uranium and its radiological characteristics so that the auditor can look for problems and make appropriate evaluations and recommendations. The auditor should not only examine the ALARA program plan and records, but should also visit the working areas and laboratories in the facility with a knowledgeable escort who can answer questions about activities and conditions in the facility. Reviews and/or audits provide the means to evaluate the effectiveness of the ALARA program through a detailed analysis of the data. Through these analyses, specific opportunities for improvement may be identified. For example, the exposure experience of a specific group can be tracked to evaluate trends and their probable causes. An increasing exposure trend can signal degradation in the radiological control program, a need for specialized training, changes in the work force, or a change in equipment or operational procedure in the areas in which higher exposures are being experienced. Similarly, a decreasing exposure trend could mean either that the ALARA program is accomplishing its objective or that a major change in radiological work has occurred. Such trends should be examined at least quarterly to permit initiation of timely corrective actions. When exposure trends and probable causes are clearly understood, the information should be provided to both management and staff. If an increasing exposure trend is identified, it can call attention to the problem allowing corrective action to be taken or to signal special procedures or precautions that may be needed. When the ALARA program is successful in reducing exposures, immediate feedback can verify program effectiveness and encourage further support of the program. Reviews and/or audits and communication of the results provide the base for program upgrade. Audits and/or reviews are also an effective means to evaluate the effectiveness of a policy or procedure change and assist in determining what changes are most effective for a given set of conditions, provide a basis for future decisions as to effective means for reducing exposure, provide a basis for comparing costs with results, and provide a measure of the program’s effectiveness for controlling individual and person- rem exposures as well as dose ranges and percentage of total person-rem represented by the ranges. 3.2.2.7 ALARA at Uranium Processing Facilities

The ALARA concept has wide application and serves as a basis for sound radiological control programs. The fundamental ALARA objective is to reduce radiation doses to the lowest practical levels commensurate with sound economics and operating practices. Realistic numerical goals can be set and achieved; however, compliance with numerical standards does not provide evidence that the ALARA concept is fully incorporated in the radiological control program. Rather, the success of a mature ALARA program is measured by many factors including intangibles, such as dedication to the concept of 3-8 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities dose control. A set of ALARA recommendations will therefore include both numerical goals and some relatively general philosophical guidance that, by itself, may not appear to assist in achieving ALARA goals. Development and implementation of an ALARA program in many uranium facilities may be a challenging task, due primarily to the fact that penetrating radiation doses are typically low and few individuals are exposed near the regulatory limits for occupational exposures. As a result, convincing management to spend valuable funds to further reduce radiation exposures can be a problem. The ALARA program must have the support and active participation of all levels of management. It must be understood by the worker in the field and receive his or her continued support and attention. Detailed guidance on developing and implementing an effective ALARA Program is provided in DOE G 441.1-2. 3.2.3 External Dosimetry Program The details of the external dosimetry program are discussed in Chapter 6 of this Technical Standard and in DOE G 441.1-4, External Dosimetry Program Guide (DOE 1999f). 3.2.4 Internal Dosimetry Program The details of the internal dosimetry program are discussed in Chapter 5 of this Technical Standard and in DOE G 441.1-3, Internal Dosimetry Program Guide (DOE 1999g). 3.2.5 Area Monitoring and Control The details of the area monitoring program are discussed in Chapters 4 and 5 of this Technical Standard and in DOE G 441.1-3, DOE G 441.1-4, and DOE G 441.1-8, Air Monitoring Guide (DOE 1999h). 3.2.5.1 Radiological Surveys and Data Trending Sections 835.401 - 835.403 of 10 CFR 835 establish requirements for radiological monitoring of areas and individuals. A program of routine, scheduled surveys should be established and followed, including surveys in areas that are not ordinarily expected to be affected by radiological hazards. The program should define minimum requirements, survey type, and frequency. Surveys should be performed at frequencies adequate to identify changes in posting required or an activity buildup and to ensure current radiological controls are appropriate. The surveys specified by this section should be considered minimum requirements; additional surveys should be conducted, recorded, and reviewed as necessary to ensure adequate personnel protection. Surveys should be performed to identify radiological area boundaries and the conditions within those boundaries, the appropriate posting of sources or areas, and the location and extent of localized radiological hazards. They should be performed and documented prior to the start of radiological work, during general work activities at times when changes in radiological conditions may occur, and following work to determine that final radiological conditions are acceptable and documented. A sufficient number of points should be surveyed to adequately assess the radiological status of the area being surveyed.

3-9 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Routine radiological surveys should be regularly conducted, recorded, and reviewed for all areas where personnel could be exposed to radiation or radioactive material throughout the site. Surveys should be performed at frequencies adequate to ensure protection of personnel. The following surveys should be considered the minimum. Additional surveys should be conducted, recorded, and reviewed as necessary to ensure personnel exposures are maintained ALARA. General radiation surveys should be performed to: a. identify and verify the boundaries of areas which must be radiologically controlled, b. verify that radiation levels in uncontrolled areas remain less than specified limits, c. determine the appropriate posting of localized higher radia tion levels, beams, or hot spots, d. ensure radiological conditions are acceptable and documented prior to, during, and at the completion of work that may cause changes in radiation levels to occur, and e. satisfy required predetermined procedure hold-points in work areas and adjacent areas, whenever operations are performed that may cause significant increases in radiation levels. The survey may be required as part of a radiological inspection step required by the work procedure. This includes areas above and below the work area as appropriate during special processing operations or cell decontamination, movement of permanent or temporary shielding, radioactive waste processing, and relocation of highly radioactive materials. Routine external radiation level surveys should be performed in the workplace at a frequency commensurate with the radiation hazard, to detect trends related to equipment, systems, environment, and work habits. Non-routine surveys of external radiation levels in the workplace should be performed: a. before initial use of a new installation, system, or equipment, or as soon as possible after a radiation source is brought into the area, b. whenever changes in procedures, equipment, or sources have occurred that may cause changes in the external radiation levels, c. after modification to a shield or changes in shield materials, d. as the basis for trend evaluation of external radiation level conditions, e. when a radiological accident has occurred or is suspected, or f. when requested by the personnel performing the activity. A sufficient number of points should be surveyed to adequately assess the radiological status of the area. Regular predetermined points may be used, but additional spot monitoring should be done to ensure all changes in dose rates are identified, recorded, and reviewed. All records of surveys should clearly identify, as a minimum: 3-10 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities a. name, signature, and employee number of the surveyor, b. survey instrument(s) model number, serial number, and calibration date, c. type(s) of radiation being monitored (e.g., neutron, gamma, etc.), d. dose rates, e. estimated doses to surveyors (from direct-reading dosimeters, if applicable), f. date and time the survey was performed, and g. locations where radioactive material is located temporarily (or is being temporarily stored) or where equipment that generates ionizing radiation is being operated. Records of the results of radiation surveys should be retained in accordance with facility policy.

Survey data should be reviewed by the facility radiological control supervisor. Significant findings should be presented to the facility manager in a timely manner. Radiological control personnel should summarize survey data in each building or area at least once a month. Significant changes or trends in area dose rates and/or radiological contamination should be noted and corrective actions assigned. The survey summary should be presented to the facility management monthly. Survey results and data summaries should be made available to the ALARA committee periodically and should be used to: a. provide a basis for evaluating potential worker exposure on a job and in ALARA preplanning, b. provide a baseline for trend analysis, investigation, and correction of unusual conditions, c. track the status of jobs (including identification of good practices) and detect departures from good operating procedures and/or the failure of radiation controls, and d. identify the origin of radiation exposures in the plant by location, system, or component. Radiological control personnel should post survey maps at the entrance to all radiological areas so personnel can be aware of radiological conditions within the area. A survey data trending program should be conducted to indicate the continuing effectiveness of existing control, to warn of deterioration of control equipment or effectiveness of operating procedures, to show long-term variations in radiation levels, and to identify and correct improper radiation work practices. See NUREG-0761, Radiation Protection Plans for Nuclear Power Reactor Licensees (NRC 1981), sections 07.B(I)(C), 09.B(4), and 09.C(I)(C). Radiological control personnel should perform trend analyses on all permanent radiological areas. At a minimum, one complete survey record should be evaluated and included in the trend analysis program for each survey required to be performed by the facility routine control program. See NUREG-0761 (NRC 1981), 07.B(I)(C), 09.B(4), and 09.C(I)(C). 3-11 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Radiological control personnel should use the facility reporting system to identify discrepancies and abnormal trends and should summarize the data review results in their monthly reports to the radiological control manager. Survey data trends should be investigated when either an upward trend occurs, causing a significant increase (10% or more), or an abrupt change in conditions occurs that cannot be directly correlated to normal activities. 3.2.5.2 Instrumentation Considerations Instrumentation performance criteria are necessary for portable, fixed, and emergency monitoring instrumentation. There are also requirements for instrument calibration and testing. General Performance Criteria for Instruments

Programs for in-plant monitoring of uranium consist mainly of airborne and surface contamination surveys and dose rate surveys. The general and specific performance criteria for the instrumentation needed to conduct these programs are described in ANSI N317-1991, Performance Criteria for Instrumentation Used for In-Plant Plutonium Monitoring (ANSI, 1980). Performance specifications are also given in ANSI N323-1993, Radiation Protection Instrumentation Test and Calibration (ANSI 1993), ANSI N42.17A, Performance Specifications for Health Physics Instrumentation - Portable Instrumentation for Use in Normal Environmental Conditions (ANSI 1988a), and ANSI N42.17C-1989, Performance Specifications for Health Physics Instrumentation - Portable Instrumentation for Use in Extreme Environmental Conditions (ANSI 1987a) for portable radiological control instrumentation and IEC Publication 325, Alpha, Beta, and Alpha-Beta Contamination Meters and Monitors (IEC 1981) for alpha and beta contamination meters and monitors. Criteria for air monitoring instrumentation are provided in ANSI/HPS N13.1-1999, Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilitie s (ANSI/HPS 1999b), IEC Publication 761-2, Equipment for Continuously Monitoring Radioactivity in Gaseous Effluents (IEC 1983), and ANSI N42.17B-1989, Performance Specifications for Health Physics Instrumentation - Occupational Airborne Radioactivity Monitoring Instrumentation (ANSI 1987b). Criticality alarm systems are discussed in ANSI/ANS 8.3-1986, Criticality Accident Alarm System (ANSI 1986a). The criteria discussed in the following sections are specified in these standards as referenced. Portable Monitoring Instruments ANSI N317 discusses several criteria related to the performance of portable monitoring instruments: a. The overall accuracy should be within ±20%, and the precision should be within ±10% at the 95% confidence level. b. The response time (i.e., the time for the instrument reading to go from zero to 90% of full scale) should be <10 seconds on the most sensitive scale and <2 seconds at readings of 100 mrem/h, 100 mR/h, and 500 dpm or greater. (This criterion is unrealistic with current neutron instrument capabilities. Response time is typically 30 to 60 seconds.) c. The instrument should be able to maintain accuracy and precision for a minimum of 24 hours of continuous operation. 3-12 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities d. The instrument should have a minimum battery lifetime of 200 hours of continuous operation. ANSI N42.17A specifications differ slightly. e. The response of the instrument should not change by more than ±15% from a reference value taken at 20°C over the anticipated temperature range for operation. f. The instrument system should function within specifications over all anticipated combinations of temperature and humidity (e.g., 15° to 65°C, 40% to 95% relative humidity). ANSI N317 states the minimum detection capability for alpha monitoring instruments ideally should be 220 dpm/100 cm2 of surface area and should not be more than 500 dpm/100 cm2. This requirement should be met in the presence of a radiation field of 0.10 rem/h of neutrons in the energy range of thermal to 10 MeV, and/or in the presence of 0.10 rem/h of photons in the energy range of 0.010 to 1.25 MeV. The operating range should be from 0 dpm to at least 100,000 dpm/100 cm2 of surface area. The response of the instrument to beta-interfering radiation is an important specification that should be stated by the manufacturer.

Photon monitoring instruments should meet the accuracy requirements stated in ANSI N317 over the energy range of 0.01 to 1.25 MeV. The angular response of this type of instrument should be within ±15% over a 2 pi steradian frontal direction using at least two photon sources with energies ranging from 0.06 to 1.25 MeV. Experience has shown this response specification is not met by most instruments at lower energies due to attenuation of the photon. The energy dependence should be within ±15% over the range of 0.01 to 1.25 MeV and the operating range should be from 0.5 mR/h to at least 5000 mR/h. Experience has shown that ±20% over 0.01 to 1.25 MeV is more realistic. This specification applies to a specific window selection (e.g., below 0.05 MeV, the electron equilibrium cap or beta shield must be removed). ANSI N42.17A has a broader scope than ANSI N317, but the criteria in it apply to portable survey instruments. Additional criteria include geotropism (maximum change of 6% from reference reading for all orientations), temperature shock, mechanical shock, vibration, and ambient pressure (maximum change of 15% from reference reading for the latter four criteria). Some differences exist between ANSI N42.17A and ANSI N317. In most cases, the criteria for ANSI N42.17A are more applicable because these criteria are based on substantial testing, which was sponsored by DOE. In ANSI N42.17A, precision is tied into a measurement level; for example, it quotes a precision of 15% at <500 cpm and 10% at >500 cpm. Also, with the advent of liquid crystal displays and other digital readouts, "response time" is defined as the time it takes for the reading to move from 10% to 90% of the equilibrium or steady-state reading. Another significant difference in the standard is the battery lifetime specification is 100 hours instead of the 200 hours mentioned in ANSI N317. For direct alpha contamination surveys, the use of audible signals (headphones or speaker) greatly facilitates the detection of "hot spots." IEC Publication 325 provides additional guidance on the uniformity of probe response for alpha and beta contamination meters. Surface sensitivity measurements are also discussed in this standard. 3-13 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Performance Criteria for Fixed Monitoring Instruments Airborne contamination monitors, surface contamination monitors, and photon area monitors, and emergency instrumentation are fixed monitoring instruments subject to the following standard performance criteria. Airborne Contamination Monitors. Airborne contamination monitors, normally CAMS should meet the following criteria according to ANSI N317. The primary purpose of any CAM is to detect the presence of airborne radioactivity and activate an alarm to warn personnel in the area so actions can be taken to minimize personnel exposures. The goal for any CAM should be to perform this function as quickly as possible and at the lowest detectable level of radioactive airborne concentration. The quantity of airborne radioactivity that will result in an alarm within a given time interval is defined in units of DAC-h for a particular radionuclide and is a function of the nuclide’s airborne concentration in DACs, the sampling rate, the lower limit of detection of the instrument, and the time needed for the alarm to occur. Mishima et al. provides guidance on each of these functions.

ANSI N42.17B provides additional performance criteria for air monitors used to detect uranium. This standard provides specifications for general criteria (sampler design, units of readout, alarm threshold, etc.), electronic criteria (alarms, stability, response time, coefficient of variation, and line noise susceptibility), radiation response, interfering responses (radiofrequency, microwave, electrostatic, and magnetic fields), environmental criteria (temperature, humidity, and pressure), and air-circuit criteria. More detailed specifications are provided in ANSI N42.17B than in ANSI N317; however, the environmental criteria and the limits of variation are not as restrictive as those in ANSI N317. With respect to accuracy, ANSI N317 requires less than ±20%, and ANSI N42.17B requires 40% at the 95% confidence level. For the environmental criteria, ANSI N317 requires that the readings change less than 5% under ambient conditions, while ANSI N42.17B gives a 15% limit of variation. As discussed previously, criteria from ANSI N42.17B are more applicable because they are supported by instrument testing. ANSI N13.1 provides detailed guidance on sampling methods from stacks and ducts. One criterion that relates to CAMs is that air sample lines between air inlet and filter media should be eliminated where possible; where not possible, they should be designed to meet the sampling criteria contained in the standard (e.g., short lines, proper sampling rate, smooth bends). The use of Tygon tubing as sample lines should be minimized or eliminated. Air in-leakage from surrounding areas can be a problem when using sampling lines. Testing for air in-leakage should be performed at least annually or when seals or "O" rings are replaced. Surface Contamination Monitors. Surface contamination monitors include hand and/or shoe counters and instruments (or probes) with sufficient flexibility to survey pieces of equipment, including exterior clothing. ANSI N317 states these instruments should have an audible alarm, a frequency that is proportional to the count rate, or a pre-selectable trip setting, and upon reaching that level, should activate an audible or visible alarm or both. These instruments should be calibrated according to the requirements in ANSI N323 and be equipped with a check source. Fixed instruments should be powered by alternating current (AC) and provided with an emergency power source. Performance Criteria for Emergency Instrumentation Meeting the criteria for criticality accident alarm systems, fixed nuclear accident dosimeters, and other emergency instrumentation is essential. 3-14 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities Criticality Accident Alarm Systems (CAAS). See section 7.0 for discussion of nuclear criticality safety, including CAAS. Fixed Nuclear Accident Dosimeters. All DOE facilities that have sufficient quantities and kinds of fissile material to potentially constitute a critical mass should provide nuclear accident dosimetry. Requirements for fixed nuclear accident dosimeters are found in 10 CFR 835.1304 and DOE Order 420.1A, Facility Safety (DOE 2002). Effluent Monitors. Facilities should evaluate potential emissions in accordance with ANSI/HPS N13.1 to determine the need for stack sampling and/or monitoring.

Other Emergency Instrumentation. Other emergency instrumentation should provide ranges for all radiation dose rates and contamination levels potentially encountered at the time of an accident. Normally, dose rate capabilities from a few millirem per hour to a few hundred rem per hour should be required while capability requirements for the contamination level may range upward from 200 dpm/100 cm2 for alpha contaminants and 100 dpm/100 cm2 for beta-gamma emitters. Performance specifications for emergency radiological monitoring instrumentation can be found in ANSI N320-1979, Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation (ANSI 1975) and BNWL-1742, Technological Consideration in Emergency Instrumentation Preparedness. Phase II-B - Emergency Radiological and Meteorological Instrumentation for Mixed Oxide Fuel Fabrication Facilities (Andersen et al. 1974). Instrument Calibrations and Testing Radiation doses and energies in the work areas should be well characterized. Calibration of instruments should be conducted where possible under conditions and with radiation energies similar to those encountered at the work stations. Knowledge of the work area radiation spectra and instrument energy response should permit the application of correction factors when it is not possible to calibrate with a source that has the same energy spectrum. All calibration sources should be traceable to recognized national standards. When the work areas have been well characterized, the calibration facility used by the uranium facility should be set up to represent as closely as possible the work area’s radiation fields. DOE G 441.1-7, Portable Monitoring Instrument Calibration Guide (DOE 1999i) and ANSI N323 provide guidance on radiation monitoring instrument calibration. The reproducibility of the instrument readings should be known prior to making calibration adjustments. This is particularly important if the instrument has failed to pass a periodic performance test (i.e., the instrument response varies by more than ±20% from a set of reference readings using a check source) or if the instrument has been repaired. The effect of energy dependence, temperature, humidity, ambient pressure, and source-to detector geometry should be known when performing the primary calibration. Primary calibration should be performed at least annually. Standards referenced in Section 3.5.2 discuss specific performance testing of radiation detection instruments. Testing procedures in these standards should be used for periodic requalification of instruments or detailed testing of instruments. The calibration of photon monitoring instruments over the energy range from a few keV to 300 keV is best accomplished with an x-ray machine and appropriate filters that provide known x-ray spectra from a few kiloelectron volts to approximately 300 keV. Radionuclide sources should be used for higher 3-15 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities energies. Most ion chambers used to measure photon radiations have a relatively flat energy response above 80 to 100 keV; 137Cs or 60Co are typically used to calibrate these instruments. These sources also should be used to calibrate Geiger-Mueller (GM) type detectors. It should be noted that some GM detectors (e.g., those with no energy compensation) can show a large energy dependence, especially below approximately 200 keV.

Whenever possible, beta detectors should be calibrated to the beta energies of interest in the workplace. A natural or depleted uranium slab source can be used for calibration of beta detectors when beta radiations in the workplace have energies similar to the uranium. International Organization for Standardization beta sources should be used for all other purposes: the energy dependence of beta detectors can be tested using the calibration sources listed in the ISO Publication 1980 (ISO, 1984); these include 90Sr, 90Y, 204Tl, and 147Pm. The calibration and testing of crucial monitoring systems are extremely important to the overall radiation protection program, but have often been neglected. Effluent monitoring and sampling systems (when present) and remote area monitoring systems should be given several tests. The radiological, environmental, and mechanical characteristics of the instrumentation portion of the system should be fully evaluated prior to its first use to ensure its compatibility with performance requirements and facility operating conditions. The effluent sampling losses from the sample probe to the collector/detector should be determined. This test should be repeated at least annually and when a significant change in the sampling equipment is made. The sample probe should be examined at least once a year to verify its design or performance has not been changed by corrosion. The recorder of the sample flow rate should be calibrated when it is installed and annually thereafter. The operability of the overall system should be completely tested once, with repeat tests only after modification, repair, or maintenance. Operability checks should be scheduled at least monthly and calibration performed at least annually. The operation of criticality or other radiation alarm signal systems should be checked periodically to ensure the alarms are audible at all potentially occupied locations (ANSI 1986a). To prevent any desensitizing of staff, the staff should be aware the tests will be performed, and where possible, tests should be scheduled during off-shift hours. Building systems should be tested semiannually and the area-wide system should be tested at least annually. Any portion of the detector/alarm system affected by the test should be reconfirmed for operability after the test is completed (e.g., if a detector is disconnected and a signal is injected at that point, the detector should be tested immediately after it has been reconnected). 3.2.6 Radiological Controls 3.2.6.1 Work Authorizations Written authorizations shall be required to control entry into and work within radiological areas and shall specify radiation protection measures commensurate with the existing and potential hazards (10 CFR 835.501(d)). ALARA considerations need to be included in the work authorization. One approach that works well is the inclusion of an ALARA worksheet with the radiological work permit (RWP). Although the written work authorizations may take any appropriate form (e.g., written procedures, policy statements, technical work documents, etc.), RWPs are most often used. RWPs should be used for entry into high and very high radiation areas, high contamination areas, and airborne radioactivity areas. RWPs should also be used to control entry into radiation and contamination areas and for handling materials with removable contamination. The RWPs should be initiated by the work group responsible 3-16

DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities for the activity. All RWPs should be reviewed and approved by the radiological control staff and cognizant line management. The RCS provides detailed guidance for RWPs. Radiological workers should read and understand the applicable RWP before entering the affected area. Copies of the RWP should be located at the access point to the applicable area. Workers should acknowledge by signature or through electronic means that they have read, understood, and will comply with the RWP before they initially enter the area and after changes. Out-of-date RWPs should be removed. 3.2.6.2 Facility Posting and Labeling Radiological areas, controlled areas, and radioactive material areas shall be posted, unless the conditions constituting the authorized exceptions specified in 10 CFR 835 exist (10 CFR 835.601-835.606). DOE Guide G-441.1-10, Posting and Labeling for Radiological Control Guide (DOE 1999j) and the RCS provide appropriate guidance. The technical criteria for defining the required areas should be established, documented, and consistently applied. The radiological control staff should establish and document the conditions that require areas to be barricaded and marked to prevent personnel from inadvertently entering them and to be physically locked to preclude unauthorized personnel from entering them. Entrance to areas where radioactive materials are used or stored should be restricted, based upon established criteria. The radiological control staff should post current surveys at the access control point for use in pre-job planning. Additional precautions, such as protective clothing, dosimetry, and respiratory protection requirements should also be posted. 3.2.6.3 Unposted Areas Certain areas of facilities that handle radioactive materials should be maintained free of detectable radioactive contamination. These areas should also be maintained at ambient radiation levels equivalent to the environmental background of the facility. Parts of the facility that should meet these requirements include lunchrooms, offices, restrooms, janitor rooms, corridors outside operational areas, foyers, and outside areas surrounding the facility, including building roofs. To determine that these areas meet the requirements of non-radioactive cleanliness, they should be surveyed with count-rate instruments sensitive to the radioactive isotopes of interest. These clean areas should be maintained below the detection levels cited in 10 CFR 835. 3.2.6.4 Visits by Regulatory Personnel Periodically, personnel from DOE and other Federal and state agencies visit radiological facilities for audit purposes or to discuss regulatory changes. In most cases, they will look at records of the radiation protection program and, in some cases, will also enter posted areas of the facility. These regulatory personnel should have ready access to the facility; provided that applicable training, dos imetry, and other requirements are met. They should have complete access to facility personnel knowledgeable in the subjects they wish to discuss. New commitments requested should be referred to the appropriate facility and DOE management. 3-17 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.2.7 Emergency Exposure Situations Requirements and guidance for emergency exposure situations are discussed in detail in Chapter 9 of this TS.

3.2.8 Nuclear Accident Dosimetry Nuclear accident dosimetry is discussed in detail in Chapter 6 of this TS. 3.2.9 Records The systematic generation and retention of records relating to the occupational radiological control program are essential to describe the occupational radiation dose received by individuals and the conditions under which the exposures occurred. Such records have potential value for medical, epidemiological, and legal purposes. 10 CFR 835 establishes radiation protection program records requirements. Detailed guidance is provided in DOE G 441.1-11, Occupational Radiation Protection Record-keeping and Reporting Guide (DOE 1999k) and the RCS provide guidance for radiation protection program records. The following types of records should be maintained: a. Individual radiological exposure records 1. internal doses, 2. external doses (whole body, skin of the whole body, extremities, and lens of the eye), 3. total effective dose equivalent (summation of internal and external doses), 4. lifetime and cumulative total effective dose equivalent, 5. non-uniform exposure to the skin, 6. supportive data for determining individual doses, and 8. individual medical records. b. Radiological status of work area records 1. radiation safety analysis and evaluation reports, 2. radiation work procedures and permits (RWPs), 3. radiation and contamination surveys, 4. records of releases of potentially contaminated materials and equipment from radiological areas, 3-18 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 5. airborne radioactivity monitoring records, and 6. area monitoring instrumentation records. c. Records of monitoring methods 1. radiation protection policies and procedures, 2. evaluation of exposure data, 3. functional capabilities of dosimeters and instruments, 4. calibration and maintenance records, 5. audits and programmatic reviews, 6. changes in procedures, techniques, and equipment, and 7. individual radiation safety training. Most of the required radiological records have established retention periods. The retention periods are discussed in DOE Guide 1324.5B, Records Management (DOE 1996b). Individual records may be covered by the Privacy Act; the DOE has codified the Privacy Act in 10 CFR 1008, Records Maintained on Individuals (Privacy Act) (DOE 1994b). 3.2.10 Radiation Safety Training A thorough radiation safety training program should be established at uranium facilities. Training programs should ensure that personnel have the training to work safely in and around radiological areas and to maintain their individual radiation exposure and the radiation exposures of others ALARA. Separate training programs should be established for general employees and radiological workers. DOE’s core training materials should form the basis for the train ing programs, and should be augmented with site- specific information. The training of all staff members should be carefully documented. DOE G 441.1-1A, DOE G 441.1-12, Radiation Safety Training Guide (DOE 1999l) and the RCS provide guidance on information to be presented during the training programs. DOE requires biennial radiation safety training for general employees and radiological workers. In the alternate year when retraining is not performed, refresher training should be provided. Individuals who work with uranium should have special uranium facilities training in addition to or as part of the appropriate level of Radiological Worker Training.

3.2.10.1 Radiological Worker Training Before working in uranium operations, all radiological workers should be trained and qualified. A thorough radiation protection training program should be established at uranium facilities. Before beginning uranium training, each uranium worker should receive General Employee Radiological Training (DOE 1998e) or either Rad Worker I or Rad Worker II Training (DOE 1998d). In addition, DOE-HDBK-1113-98 Radiological Safety Training for Uranium Facilities (DOE 1998c) provides guidance on providing radiation safety training to workers at uranium facilities. 3-19 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities The level of radiation worker training should be determined in accordance with the Table 3-1 of the RCS. All training should be consistent with the guidance provided in DOE G 441.1-12. All training dispositions and records should be documented in accordance with 10 CFR 835.704. 3.2.10.2 Training for Other Facility Personnel Non-radiological workers in a uranium facility should be given a general orientation on the radiation safety concerns for working with uranium, the general protective measures used for work with uranium, and the engineered safety features of the facility. 3.2.10.3 Members of the Public Members of the public with a demonstrated need to enter the following areas may be allowed access if such access is controlled with a combination of training and the use of escorts trained for the specific area: a. Radiological Buffer Areas, b. Radiation and High Radiation Areas, c. Contamination Areas, and d. Radioactive Material Areas. Guidance for training of members of the public is provided in DOE G 441.1-12 and the RCS. Individuals under 18 years of age should not be permitted to enter radiation areas or contamination areas without the approval of the radiological control manager. Area entry requirements and access restrictions for members of the public should be established in facility procedures. Members of the public should be prevented from entering very high radiation, high contamination, and airborne radioactivity areas. All facility personnel serving as a qualified escort should ensure that each visitor under his/her cognizance completes a facility radiological visitor form. The qualified escort should also sign the visitor form and complete it as appropriate. Facility-sponsored visitors should provide the following before entering radiological areas, unless these records have already been entered into the facility entry control system: 1. evidence of completing required training, as applicable 2. visitor radiation exposure disclosure 3. a medical disclosure form or the results of a medical evaluation. The host facility manager should forward the visitor radiation exposure and medical disclosure forms to Dosimetry. The use of offsite mask fit certification is authorized under the following conditions: 3-20 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 1. A mask fit has been completed within the previous year. 2. The individual presenting the mask fit certification card has not changed physical appearance in a way that would affect the seal of the mask to the face. For example, this could be determined by a combination of: review of photograph on mask fit certification card (if available), examination of facial hair on areas which could affect mask seal, and discussion with wearer of any physical changes which could affect mask seal.

3. The facility has the masks available that the individual is certified to wear. If there are members of the public who live or work near a uranium facility, a plan for orientation of members of the public should be developed to inform them of facility activities. Such a plan should include information on the concerns that require protection of people from potential injuries by uranium, the general protective measures used at the facility to confine it and keep it out of the public domain, and solicitation of information on the concerns of members of the local public about uranium. To the extent possible, efforts should be made to allay those concerns. The information in the public education plan should also be provided to local news media. 3.3 RELATED PROGRAMS 3.3.1 Onsite Packaging and Transportation The hazardous materials organization conducts onsite radioactive shipments with the assistance of radiological control. This program requires the hazardous materials organization representatives to review onsite radioactive shipping records, document the errors or omissions observed, and evaluate trends and revise training as needed. Serious deficiencies are to be documented and the reports should be submitted in accordance with DOE O 460.1B, Packaging and Transportation Safety (DOE 2003b). The packaging organization is responsible for coordinating onsite package design and prepara- tion of safety analysis documentation. The following sections describe typical process, review, and approval requirements for onsite safety analysis documentation. 3.3.1.1 Initiation New safety analysis documentation or reviews/changes to existing documentation can be requested by a user organization based on programmatic or operational requirements. The request is submitted in writing to the packaging organization and includes proper justification and support documentation. The packaging organization makes routine revisions as necessary to reflect policy and regulation changes. 3.3.1.2 Preparation The packaging organization coordinates the analysis, prepares safety analysis documentation, and guides the documentation through the review and approval process, including the resolution of review comments and the obtaining of required approval. 3-21 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.3.1.3 Control Safety analysis documentation is prepared and maintained according to facility policy. The document control system provides an accessible, auditable, and retrievable method for maintaining and changing safety analytic documentation. 3.3.1.4 Review and Approval Cycle Safety analysis documentation is reviewed, approved, and changed according to facility policy. Additional reviews and approvals include the following people and organizations: a. user, b. cognizant engineer, c. packaging organization, d. quality assurance, e. responsible environmental assurance organization, onsite only, f. packaging, shipping, and waste safety assurance organization, g. criticality engineering analysis, if criticality analysis is required, h. packaging and shipping approval authority, and i. DOE field office, if the package is to be used for HRCQ inter-area shipments. 3.3.1.5 Approval for Editorial Changes Inconsequential editorial changes to a safety analysis document may be approved at the operating level. 3.3.1.6 Utilization

Once a safety analysis document is approved, copies are sent to the affected organizations, including operations and applicable facility engineering, to incorporate the administrative controls from the safety analysis document into the affected operating documents. User organizations must obtain the packaging organization review of all operating procedures that incorporate instructions or administrative controls found in COCS, SARPS, SEPS, DAPS, DOT exemptions, and Federal and state packaging requirements to ensure that they are properly incorporated. Onsite packages currently approved for onsite use should be cataloged and described in a hazardous materials packaging directory maintained by the packaging organization. New packages are added to the directory as they are developed and approved. 3-22 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.3.2 Conduct of Operations The organization and administration of operations should ensure a high level of performance in DOE facility operations is achieved through effective implementation and control activities. Administration of operations activities should recognize that protection of the environment, maintaining a high-quality safety program and productivity are compatible goals. DOE policies and standards describe the standards of excellence under which the facility is expected to operate. Clear lines of responsibility for normal and emergency conditions must be established. Effective implementation and control of operating activities are achieved primarily by having readily accessible written standards for operations, periodical monitoring and assessment of performance, and personnel accountability for performance. For a more detailed discussion, see DOE Order 5480.19, Ch.2, Conduct of Operations Requirements for DOE Facilities (DOE 2001a). A high level of performance in DOE operations is accomplished by management establishing high operating standards and then by communicating the operating standards to workers by providing sufficient resources to the operations department, ensuring personnel are well trained by closely monitoring performance in operations, and holding workers and their supervisors accountable for their performance in conducting activities. Senior management establishes operating standards, considering input from workers when appropriate. Working-level personnel will more strongly support the standards when they have had appropriate input into their development. Standards should define operating objectives, establish expected performance levels, and clearly define responsibilities in plant operations. Standards for operating activities should be integrated into operations department procedures and programs. Operating standards should also be communicated to workers by training them in operating practices and by having supervisors monitor and guide work involving facility operations. Sufficient staff, facilities, equipment, and funding should be allocated to permit the operations department to effectively perform its functions. Performance in operations should be closely monitored by facility management, preferably using operating reports and goals, so the performance of the operations department can be effectively measured. Operations personnel should be held accountable for their performance through supervisor counseling, performance appraisals, and, when necessary, disciplinary measures. Remedial training should be provided when appropriate.

The radiological control organization, as a support element, must ensure that all aspects of radiation safety are considered in the establishment of operations standards and policy. A well-instituted cooperative relationship between operations and radiological control is paramount to the health and safety of workers and the public and to protection of the environment. A uranium facility should have a written policy on radiation protection, including an ALARA policy. All radiation protection procedures and controls should have recognizable or formal technical bases for limits, methods, and personnel protection standards. Procedures should be adequately documented, updated periodically, and maintained in a centralized historical file. A control system should be established to assure all copies are accounted for and all new procedures are included in the historical files. A designated period of time for holding the historical files should be established. ANSI N13.6-1989 (ANSI, 1966) provides guidance on historical files. In addition, radiological control procedures should have a documented approval system and established intervals for review and/or revision. A tracking system should be developed to ensure the required reviews and revisions occur. 3-23 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 3.3.2.1 Radiological Work Procedures Radiological work procedures, including RWPs, survey procedures, ALARA reviews, sample counting, and other task procedures, fall within the requirements for conduct of operations. All sections of DOE Order 5480.19 apply. The guidance and requirements of Section XVI, "Operations Procedures," is especially pertinent to radiological work procedures. Procedures are a key factor affecting radiation protection performance. Appropriate attention should be given to writing, reviewing, approving, and monitoring implementation of radiation protection procedures. There should be documented qualification and training requirements for those who prepare and approve procedures. A formal approval process should be established. Procedure changes and revisions should be subject to the same review and approval process as the initial procedure. Personnel should be trained in the use of the procedures they will be expected to perform. For RWPs, workers are required to read the RWP and verify by signature they have read it, understand its contents, and will comply with its requirements in the conduct of the work. Procedures should be available for personnel use. The RWPs should be posted at the entrance to the work location. There should be a system in place to assure posted copies of all work procedures, including RWPs, are current. 3.3.2.2 Posting and Labeling The requirements for area posting and radioactive material labeling are established in 10 CFR 835, Subpart G. Guidance on implementing the regulatory requirements can be found in DOE G 441.1-10 and the RCS. Conformance to conduct of operations requirements should assure a reasonable degree of uniformity in the posting and the signs used, as well as verifying that operator aids and other posted information do not interfere with necessary radiological posting. Radiological postings should be reviewed in the same manner as the posting of operating aids, in conformance with DOE Order 5480.19. 3.3.2.3 Instrument Calibration The status of installed and portable radiological instruments should be well known and appropriate to the use.

"Ownership" of installed monitoring instruments should be well known and the responsibility and authority for calibration, repair, and notification clearly established. Because such information is often used by more than one group, formal notification procedures should be established to cover those times when the instrument is out of service or beyond the required calibration schedule. Configuration control and quality assurance requirements for installed systems should be established commensurate with their safety significance. For portable instruments, conduct of operations requirements are normally built into the routine calibration and survey program. Functional checks are routinely made to verify calibration, instruments are checked to assure they are within the calibration period, and survey procedures require identification of the instruments used so if a problem is later found, measurements can be repeated. 3.3.2.4 Audits Conduct of operations does not, in itself, address requirements for auditing. The guidance does state that inspections, audits, reviews, investigations, and self-assessments are part of the checks and balances needed in an operating program. Auditing is one of the many tools line management has at its 3-24 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities disposal to identify problems. Each one of the 18 topics addressed in DOE Order 5480.19 should be subject to both internal self-assessment and external auditing to assure effective implementation of requirements. Any deficiencies identified should be documented and corrective actions aggressively pursued and tracked to completion. The self-assessment and audit process should include conducting trend analyses and root cause evaluations of deficiencies and communication of results throughout the organization. 3.3.2.5 Decommissioning of Weapons and Weapon Facilities Decommissioning of nuclear weapons and nuclear facilities is subject to the same conduct of operations requirements as operating facilities. In general, some components, once they are separated, can be downgraded in safety significance. Also, facilities undergoing decommissioning will have fewer safety systems. During decommissioning, status control and shift turnover are extremely important considerations. Posting of radiological areas and labeling of radioactive materials are also an increasing challenge because of the rapidly changing radiological status. In extreme cases, it may be desirable to have workers review or sign the RWP each day to ensure they are aware of the status. 3.3.3 Integrated Safety Management The radiological control program should be developed and implemented in a manner that is consistent with the DOE approved Radiation Protection Program required by 10 CFR 835.101 and the requirements of DOE Policy P 450.4, Safety Management System Policy (DOE 1996a), and its associated guidance documents. The RPP should describe a system of radiological controls that can be implemented on a site- wide basis and tailored to meet facility-and hazard-specific needs. The program should provide for increasing worker involvement in identification and implementation of appropriate controls. Like the ALARA process, an effective integrated safety management system emphasizes the development and implementation of controls that are commensurate with the hazards associated with any specified activity. Under ISM, both DOE and DOE-contractor line managers are charged with responsibility for integrating safety measures into all facets of work planning and execution. Line managers at uranium facilities should use the RCS and this TS as a guide to integrating radiological control measures into work planning and execution.

3-25 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities This page intentionally left blank. 3-26 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities 4.0 CONTAMINATION CONTROL Contamination control is an important part of the overall radiological control program. There are three main aspects to this: 1) control of the release of contamination into the work-place environment; 2) control of personnel exposure to the contamination that does get into the work place; and 3) protection of personnel from intake of contaminants. Effective control of personnel exposure to uranium and its decay products is accomplished mainly by controlling the potential for inhalation and ingestion of radioactive materials. Monitoring provides an indication of the effectiveness of physical design features and administrative controls in controlling exposure to radioactive material. This chapter addresses the basic features of an effective contamination control program and the technical considerations of implementing the program. A release of radioactive material from containment typically results in surface contamination and airborne dispersion. Airborne contaminants are continuously cleared from the work place by ventilation. Strategic air sampling detects the release of an airborne contaminant and provides the means for control, minimization of personnel exposure, and evaluation of inhalation exposure. Considerations for design of an air monitoring program are followed in this chapter by a section on surface contamination control. Finally, protection of personnel from contaminant intake is accomplished with protective clothing and respiratory protection. 4.1 AIR MONITORING The most common route of uranium intake for workers is by inhalation. Airborne particles deposit throughout the respiratory tract. Some of the deposited particles are swallowed, contributing to ingestion, requiring that both inhalation and ingestion be considered with an exposure to airborne material. The particle size distribution that determines deposition in the respiratory tract is affected by the mechanism of dispersion and the nature of the source material. Characterization of inhalation exposure should make use of all available information about the chemical and physical form of airborne material. This information, along with spatial and temporal distribution, provides the basis to minimize personnel exposure for air contamination control. 4.1.1 Internal Versus External Dose Philosophy The widespread application of methods to contain uranium in DOE facilities has resulted in a history of relatively minor internal exposures. The methods used to control internal dose have been developed for a varie ty of reasons: a. The assessment of internal dose requiring bioassay is difficult, imprecise, time-consuming, and offensive to personnel as compared to external dosimetry. For example, an accidental internal uptake may require the subject to submit dozens of biological samples over the span of many months, as well as requiring extensive analytical support for measurement of sample content, considerable time of trained professionals to analyze data and calculate the internal dose, and long lapses before dose estimates are available, thus handicapping the assessment of the occupational exposure status and treatment of the worker. 4- 1

DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities b. Prevention of internal exposure is often more feasible and successful than prevention of external exposure. Contained radioactive material may continue to produce external penetrating fields of radiation, but no internal exposure potential. Portable protective devices (respiratory protection equipment) can minimize internal exposure when containment is not practical. c. Recommendations of the ICRP in formulating a dose equivalent limit system have resulted in combining internal and external dose. Again, the difficulty and time delay of internal dosimetry make elimination of significant internal exposure an economic incentive. In facilities that process large quantities of uranium, however, there may be situations in which exposure to work-place airborne activity at low levels occurs daily. The fact that tons of material are handled, rather than gram quantities, and that the material is less toxic (on a mass basis because of low specific activity), make total containment impractical. 4.1.2 Purpose of Air Monitoring The goal of the air monitoring program is to identify, evaluate, and control internal dose received by workers from routine occupational exposure to airborne radioactive materials, to confirm that source controls are functioning properly, and to assess the exposure resulting from an unusual event. There are two general aspects of air sampling that must receive equal consideration in a properly executed monitoring program. The first involves the methods and equipment by which a sample is collected and analyzed to yield an accurate measurement of the specific radionuclides. The second is the protocol of sampling location, duration, and frequency that focuses on determination of the radionuclide exposure in the work area. Air monitoring should include both active and passive air samplers. A continuous air monitor (CAM) provides for immediate alarm, warning workers of an unusual release of high levels of airborne radioactive material. This active monitoring is needed for high hazard and high potential areas to provide immediate and timely protective response, while passive sampling provides high-sensitivity activity records, trends, continuous documentation, etc. Three types of air samplers are used to accomplish the air monitoring: general area sampling (GAS), breathing zone sampling (BZS), and personal air sampling (PAS). The CAM continuously draws air through a sampler that has an active radiation detector. The sampled air is automatically monitored for an increase above normal or background levels of contamination. When airborne activity exceeds the alarm level, workers are warned of the potential problem and prompted to follow alarm procedures. This type of monitor is usually practical only for stationary samplers (GAS or BZS). It is important that a CAM be placed to sample air that accurately represents the most likely area of material release. This will protect most workers from a worst-case exposure and minimize total work-force exposure. 4- 2 DOE-STD-1136-2004 Guide of Good Practices for Occupational Radiological Protection in Uranium Facilities General Air Samplers (GAS) Air sampling is performed at a single point in the general area of a site where work with radioactive

material is being performed. The sampler is placed in a position to give the best overall representation of the area, often in the main airflow exiting the area. Airflow patterns can be determined by tests with tracer smoke or balloons. This method is typically used to measure airborne radioactivity for the following purposes: a. to determine if the work-place environments are free of significant contamination and are inherently safe for routine occupational activities, b. to detect measurable air activity which would signal the need for use of respiratory protection equipment, c. to detect unexpected loss of containment or malfunction of systems (which may not be detected by a CAM), and provide the basis to initiate corrective actions, d. to detect low-level trends in activity which can signal a gradual loss of containment in early stages, and e. to estimate personnel exposure retrospectively and evaluate compliance with applicable requirements. Breathing Zone Samplers (BZS) Breathing zone sampling is performed by placing air samplers in the immediate area in which workers will spend the majority of their time. The intent is to measure the air activity concentrations to which the workers are actually exposed. The purposes of breathing zone sampling are the same as those listed for general air sampling, but involve a greater number of samples, which gives more realistic information. Breathing zone samplers give earlier, more sensitive detection of release from containment. Samples should be collected on a schedule corresponding to individual worker activities to best represent inhalation exposure. GAS is generally not a good measurement with which to estimate internal dose. A well-placed network of BZS gives a better representation of inhalation exposure. Personal Air Samplers (PAS) Personal air sampling should give the most realistic measurement of individual worker exposure. This involves greater expense, however, to equip personnel with samplers and to process all of the individual samples. Personal air sampling is performed with a small, batte