DOE-STD-1136-2017, Good Practices for Occupational Radiological Protection in Uranium Facilities
Functional areas: Worker Protection, Lessons Learned, Radiation Protection, Radiation Doses
This Technical Standard provides operational guidance, practical lessons learned and experience gained. guides to good practice, and reference information on safe handling of uranium. Supersedes DOE-STD-1136-2009.
Supersedes:
Superseded By:
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Superseded by
A newer version replaces this document.
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Section 1
NOT MEASUREMENT
SENSITIVE
DOE‐STD‐1136‐2017
DOE STANDARD
GOOD PRACTICES FOR
OCCUPATIONAL RADIOLOGICAL
PROTECTION IN URANIUM FACILITIES
U.S. Department of Energy AREA SAFT
Washington, DC. 20585
DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited.
DOE‐STD‐1136‐2017
Available to the public on the DOE Technical Standards Program Web Site at
http://energy.gov/ehss/services/nuclear‐safety/department‐energy‐technical‐standardsprogram.
ii
http://energy.gov/ehss/services/nuclear-safety/department-energy-technical-standardsprogram
DOE‐STD‐1136‐2017
FOREWORD
This Technical Standard 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 Technical
Standard provides guides to good practice, updates existing reference material, and discusses 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 Technical Standard provides
information to assist uranium facilities in complying with Title 10 of the Code of Federal Regulations,
Part 835 (10 CFR Part 835), Occupational Radiation Protection and various DOE Orders. This technical
standard supplements DOE G 441.1‐1C, Radiation Protection Programs Guide for Use with Title 10, Code
of Federal Regulations, Part 835, Occupational Radiation Protection (DOE, 2008a) and DOE‐STD‐1098‐
2008, Radiological Control (DOE, 2009c).
This Technical Standard has been updated to include provisions in the 2007 amendment to 10 CFR Part
835. This amendment updated the dosimetric terms and models for assessing radiation doses, both
internal and external. Of particular interest for this Standard, the biological transportability of material is
now classified in terms of absorption types: F (fast), M (medium) and S (slow). Previously this was
classified in terms of material class: D (days), W (weeks) and Y (years). Throughout this Standard,
discussions of previous studies describing the biological transportation of material in the body will
continue to use D, W and Y, as appropriate. Discussions of other requirements which have not amended
their dosimetric terms and models continue to use the older terminology.
This Technical Standard does not include every requirement applicable to DOE uranium facilities.
Section 2
Individuals responsible for developing and implementing radiation protection programs at uranium
facilities should be knowledgeable of the requirements that apply to their facilities.
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DOE‐STD‐1136‐2017
Table of Contents
1 INTRODUCTION..................................................................................................................................... 1
1.1 Purpose and Applicability ............................................................................................................. 1
1.2 Definitions ..................................................................................................................................... 1
1.3 Discussion...................................................................................................................................... 1
2 PROPERTIES AND RELATIVE HAZARDS .................................................................................................. 2
2.1 Nuclear Properties of Uranium..................................................................................................... 2
2.1.1 Isotopic Characterization ...................................................................................................... 2
2.1.2 Decay Chains ......................................................................................................................... 5
2.1.3 Enrichment............................................................................................................................ 8
2.1.4 Contaminants from Recycled Uranium and Associated Hazards.......................................... 9
2.2 Physical and Chemical Properties ............................................................................................... 14
2.2.1 Uranium Fuel Processing..................................................................................................... 14
2.2.2 Uranium Metal .................................................................................................................... 15
2.3 Radiological Characteristics and Effects ..................................................................................... 16
2.3.1 Alpha‐Neutron External Hazard.......................................................................................... 17
2.3.2 Mode of Uranium Entry into the Body ............................................................................... 17
2.4 Chemical Toxicity ........................................................................................................................ 19
2.4.1 Human Response Indicators ............................................................................................... 23
2.4.2 Transfer to the Fetus........................................................................................................... 24
2.5 Chemical versus Radiological Hazards ........................................................................................ 24
Section 3
2.6 Natural Uranium Balance in Man................................................................................................ 31
2.7 Industrial Hazards ....................................................................................................................... 32
2.7.1 Hydrogen Fluoride .............................................................................................................. 32
2.7.2 Nitric Compounds ............................................................................................................... 33
2.7.3 Hydrogen Gas...................................................................................................................... 33
2.7.4 Fire ......................................................................................................................................33
3 RADIATION PROTECTION .................................................................................................................... 35
3.1 Regulations and Standards ......................................................................................................... 35
3.2 Radiation Protection Programs................................................................................................... 35
3.2.1 Organization and Administration........................................................................................ 36
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DOE‐STD‐1136‐2017
3.2.2 ALARA Program................................................................................................................... 41
3.2.3 External Dosimetry Program............................................................................................... 45
3.2.4 Internal Dosimetry Program ............................................................................................... 45
3.2.5 Area Monitoring and Control.............................................................................................. 45
3.2.6 Radiological Controls .......................................................................................................... 54
3.2.7 Emergency Exposure Situations.......................................................................................... 55
3.2.8 Nuclear Accident Dosimetry ............................................................................................... 55
3.2.9 Records................................................................................................................................55
3.2.10 Radiation Safety Training .................................................................................................... 56
3.3 Related Programs........................................................................................................................ 58
3.3.1 Onsite Packaging and Transportation................................................................................. 58
3.3.2 Approval for Editorial Changes ........................................................................................... 59
Section 4
3.3.3 Conduct of Operations........................................................................................................ 59
3.3.4 Integrated Safety Management.......................................................................................... 62
4 CONTAMINATION CONTROL............................................................................................................... 64
4.1 Air Monitoring............................................................................................................................. 64
4.1.1 Internal Versus External Dose Philosophy .......................................................................... 64
4.1.2 Purpose of Air Monitoring .................................................................................................. 65
4.1.3 Regulations and Limits ........................................................................................................ 67
4.1.4 Theoretical Considerations and Uncertainties.................................................................... 67
4.1.5 Samplers and Instrumentation ........................................................................................... 72
4.1.6 Sample Activity Measurement............................................................................................ 76
4.1.7 Continuous Air Monitors (CAM).......................................................................................... 77
4.1.8 Monitoring Strategies and Protocols .................................................................................. 77
4.2 Surface Contamination Control .................................................................................................. 80
4.2.1 Reporting and Documenting Contamination Levels ........................................................... 81
4.2.2 Monitoring .......................................................................................................................... 82
4.2.3 Release Criteria ................................................................................................................... 88
4.2.4 ALARA Guidelines................................................................................................................ 93
4.3 Personnel Contamination Control .............................................................................................. 93
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DOE‐STD‐1136‐2017
4.3.1 Monitoring Philosophy........................................................................................................ 93
4.3.2 Monitoring Program ........................................................................................................... 94
4.3.3 Protective Clothing.............................................................................................................. 95
4.3.4 Respiratory Protection ........................................................................................................ 95
Section 5
4.3.5 ALARA Guidelines................................................................................................................ 95
4.3.6 Release Criteria ................................................................................................................... 96
4.4 Decontamination and Decommissioning Techniques................................................................. 96
4.4.1 Personnel Decontamination ............................................................................................... 96
4.4.2 Equipment and Surface Decontamination.......................................................................... 97
5 INTERNAL DOSIMETRY ........................................................................................................................ 99
5.1 Internal Dose Evaluation Program .............................................................................................. 99
5.1.1 Performance Capabilities for Internal Exposure Monitoring............................................ 100
5.1.2 Protection of the Embryo/Fetus, Minors, and Members of the Public ............................ 108
5.2 Characterization of Internal Hazards ........................................................................................ 108
5.3 Scope of Bioassay Program....................................................................................................... 110
5.3.1 Classification of Bioassay Measurements ......................................................................... 110
5.3.2 Monitoring Requirements and Selection of Employees ................................................... 112
5.3.3 Selection of Bioassay Monitoring Techniques .................................................................. 113
5.4 Establishing Bioassay Frequency............................................................................................... 115
5.4.1 Frequency Based on Program Sensitivity.......................................................................... 116
5.4.2 Frequency Based on Potential Risk of Intake.................................................................... 118
5.4.3 Special Bioassay as Supplements to Routine Bioassay Programs..................................... 118
5.4.4 Long‐term Follow‐up Bioassay Programs ......................................................................... 119
5.4.5 Other Frequency Situations .............................................................................................. 119
5.5 Administration of a Bioassay Program...................................................................................... 119
5.5.1 In Vivo Monitoring ............................................................................................................ 121
5.5.2 Urine Sampling.................................................................................................................. 122
5.5.3 Fecal Sampling .................................................................................................................. 124
5.5.4 Conditions for Adjustments of Action Levels.................................................................... 125
5.6 Interpretation of Bioassay Results ............................................................................................ 125
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DOE‐STD‐1136‐2017
5.6.1 In Vivo Count Results ........................................................................................................ 126
5.6.2 Urine Sample Results ........................................................................................................ 127
5.6.3 Fecal Sample Results ......................................................................................................... 128
5.6.4 Use of Air Sample Data in Internal Dosimetry .................................................................. 128
5.7 Dose Assessment ...................................................................................................................... 129
5.7.1 Methods of Estimating Intake........................................................................................... 130
5.7.2 Alternate Methods of Intake Assessment ........................................................................ 131
5.7.3 Estimating Effective Dose from Intakes of Uranium......................................................... 131
5.8 Reference and Action Levels..................................................................................................... 132
5.9 Response to Suspected Intakes ................................................................................................ 135
5.9.1 Emergency Action Planning .............................................................................................. 137
5.9.2 Medical Emergency Response Plan .................................................................................. 137
5.9.3 Responsibilities for Management of Internal Contamination .......................................... 138
5.9.4 Immediate Medical Care................................................................................................... 138
5.9.5 Contaminated Wounds ..................................................................................................... 139
6 EXTERNAL DOSIMETRY...................................................................................................................... 140
6.1 DOSE LIMITS.............................................................................................................................. 140
6.1.1 Limiting Quantities ............................................................................................................ 140
6.1.2 Operational Quantities...................................................................................................... 144
6.2 RADIATIONS IN URANIUM FACILITIES....................................................................................... 144
6.2.1 Alpha and Beta Doses ....................................................................................................... 148
6.2.2 Gamma Doses ................................................................................................................... 149
6.2.3 Neutron Dose Equivalents ................................................................................................ 149
6.3 RADIATION DETECTION AND EVALUATION .............................................................................. 152
Section 7
6.3.1 Portable Survey Instruments‐‐Beta Radiation Response.................................................. 152
6.3.2 Portable Survey Instruments‐‐Gamma Radiation Response ............................................ 156
6.3.3 Portable Survey Instruments‐‐Neutron Response............................................................ 158
6.4 PERSONNEL DOSIMETRY........................................................................................................... 159
6.4.1 Energy Dependence .......................................................................................................... 160
6.4.2 Angular Dependence......................................................................................................... 160
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DOE‐STD‐1136‐2017
6.4.3 Dosimetry Practices .......................................................................................................... 160
6.4.4 Extremity Dosimetry ......................................................................................................... 161
6.4.5 Dose to Lens of Eye ........................................................................................................... 163
6.5 External Dose Control ............................................................................................................... 163
6.5.1 Time ..................................................................................................................................163
6.5.2 Distance............................................................................................................................. 164
6.5.3 Shielding............................................................................................................................ 164
6.5.4 Geometry .......................................................................................................................... 166
6.6 Recordkeeping .......................................................................................................................... 166
7 NUCLEAR CRITICALITY SAFETY .......................................................................................................... 167
7.1 REGULATIONS AND STANDARDS .............................................................................................. 167
7.2 Criticality Control Factors.......................................................................................................... 168
7.2.1 Controllable Factors .......................................................................................................... 168
7.2.2 Administrative Practices ................................................................................................... 170
7.3 CRITICALITY ACCIDENT EXPERIENCE ......................................................................................... 173
7.3.1 Types of Criticality Accidents ............................................................................................ 173
7.3.2 Summary of Past Criticality Accidents .............................................................................. 174
Section 8
7.4 CRITICALITY ALARMS AND NUCLEAR ACCIDENT DOSIMETRY................................................... 174
7.4.1 Criticality Accident Alarm System (CAAS) ......................................................................... 174
7.4.2 Nuclear Accident Dosimetry ............................................................................................. 175
7.5 RESPONSIBILITIES OF RADIOLOGICAL CONTROL STAFF............................................................ 179
7.5.1 Routine Operations........................................................................................................... 179
7.5.2 Emergency Response Actions ........................................................................................... 179
7.5.3 Special Considerations during Decommissioning Activities.............................................. 180
8 WASTE MANAGEMENT ..................................................................................................................... 181
8.1 POTENTIALLY CONTAMINATED WASTES .................................................................................. 181
8.1.1 Solid Waste ....................................................................................................................... 182
8.1.2 Liquid Waste ..................................................................................................................... 182
8.2 DESIGN OF WASTE PROCESSING SYSTEMS ............................................................................... 182
8.2.1 Objectives.......................................................................................................................... 182
8.2.2 Effluents ............................................................................................................................ 183
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DOE‐STD‐1136‐2017
8.3 Treatment .................................................................................................................................183
8.3.1 Airborne Wastes ............................................................................................................... 183
8.3.2 Liquid Waste ..................................................................................................................... 184
8.3.3 Solid Waste ....................................................................................................................... 184
8.4 Monitoring ................................................................................................................................185
8.4.1 Air and Gaseous Effluents ................................................................................................. 185
8.4.2 Liquid Effluents ................................................................................................................. 185
8.4.3 Water Collection System................................................................................................... 186
8.5 Waste Minimization.................................................................................................................. 186
Section 9
9 EMERGENCY MANAGEMENT ............................................................................................................ 189
9.1 Emergency Management in DOE .............................................................................................. 189
9.1.1 Basis for DOE Emergency Management Policy ................................................................. 189
9.1.2 Requirements Pertaining to All DOE Operations .............................................................. 190
9.2 Specific Guidance on Emergency Management for Uranium Facilities.................................... 191
9.2.1 Technical Planning Basis ................................................................................................... 191
9.2.2 Program Elements............................................................................................................. 195
10 DECOMISSIONING......................................................................................................................... 200
10.1 REGULATORY FRAMEWORK...................................................................................................... 200
10.1.1 Resource Conservation and Recovery Act of 1976 (RCRA) ............................................... 200
10.1.2 Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) ...200
10.1.3 National Environmental Policy Act ................................................................................... 201
10.1.4 DOE Directives and Standards .......................................................................................... 201
10.2 Residual Radioactivity Levels .................................................................................................... 202
10.2.1 Public Dose Limits ............................................................................................................. 202
10.2.2 ALARA process requirements............................................................................................ 202
10.2.3 Dose Constraints ............................................................................................................... 203
10.2.4 Authorized Limits .............................................................................................................. 203
10.2.5 Survey and Monitoring ..................................................................................................... 204
10.3 Design Features for New Facilities............................................................................................ 204
10.3.1 Building Materials ............................................................................................................. 205
10.4 Ventilation Systems .................................................................................................................. 206
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Section 10
DOE‐STD‐1136‐2017
10.4.1 Piping Systems .................................................................................................................. 206
10.4.2 Soil‐Contamination Considerations .................................................................................. 206
10.4.3 Other Features .................................................................................................................. 206
10.5 DECOMMISSIONING PROGRAM REQUIREMENTS .................................................................... 207
10.5.1 Pre‐Operational and Operational Activities...................................................................... 208
10.5.2 Post‐Operational Activities ............................................................................................... 208
10.5.3 Decommissioning Activities .............................................................................................. 208
10.5.4 Post‐Decommissioning Activities ...................................................................................... 214
10.5.5 Quality Assurance ............................................................................................................. 214
10.6 Decontamination and Decommissioning Experience ............................................................... 214
APPENDIX A ‐ REFERENCES ........................................................................................................................... 1
APPENDIX B ‐ GLOSSARY ............................................................................................................................... 1
Table of Tables
Table 2‐1. Typical Isotopic Abundances (g of Isotope per 100g of Material) ........................................ 3
Table 2‐2. Properties of Radionuclides that may be found at Uranium Facilities(a) .............................. 4
Table 2‐3. Uranium Specific Activities ................................................................................................... 8
Table 2‐4. DACs for Uranium and Selected Contaminants in Recycled Uranium................................ 11
Table 2‐5. Beta Surface Dose Rates from Equilibrium Thickness of Uranium Metal and
Compounds(a) ..................................................................................................................................................................................... 16
Table 2‐6. Health Effects from Acute Intake of Soluble Uranium(a) ..................................................... 20
Table 2‐7. Toxicological Limits on Airborne Concentrations of Transportable (soluble) Uranium......21
Table 2‐8. Uranium Levels for Various Effects ..................................................................................... 22
Table 2‐9. 2005 ACGIH Threshold Limit Values (TLVs) for Selected Metals ........................................ 25
Table 2‐10. Inhalation Classification for Some Uranium Compounds ................................................... 26
Table 2‐11. Dose Coefficients for Determining Enrichments above which Radiological Hazards ............
Become Limiting.................................................................................................................. 27
Table 2‐12. Impact of Monitoring at 100 Millirem Enrichments above which Radiological .....................
Section 11
Limits Predominate (Calculations Not Shown) .................................................................. 31
Table 2‐13. Uranium Balance for Reference Man(a)............................................................................... 32
Table 4‐1. 10 CFR § 835 Appendix D Surface Contamination Values(1) (dpm/100cm2) ....................... 87
Table 5‐1. Urine Bioassay Goals(a) for 238U ......................................................................................... 102
Table 5‐2. Fecal bioassay Goals(a) for 238U.......................................................................................... 104
Table 5‐3. In Vivo Measurement Bioassay Goals for 238U(a) ............................................................... 106
Table 5‐4. Minimum Uranium Bioassay Monitoring(a,b) ..................................................................... 113
Table 5‐5. Categories and Performance Criteria for Uranium Bioassay ............................................ 117
Table 5‐6. Minimum Suggested Frequencies for Routine Bioassay for Uranium(a) ........................... 118
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DOE‐STD‐1136‐2017
Table 5‐7. Uranium Levels for Internal Dosimetry Notification......................................................... 133
Table 5‐8. Uranium Contamination Levels for Notification of Occupational Medicine Physician.....133
Table 5‐9. Early Bioassay Measurement Results Corresponding to the Therapeutic Intervention
Action Levels Used at the Hanford Site (Carbaugh et al., 1995)....................................... 136
Table 6‐1. Effective Depth of Tissue for Various Organs ................................................................... 140
Table 6‐2. Radiation Weighting Factors, wR....................................................................................... 141
Table 6‐3. Tissue Weighting Factors .................................................................................................. 142
Table 6‐4. Radiation Dose Limits for DOE and DOE Contractors ....................................................... 143
Table 6‐5. Spontaneous Fission Neutron Yields................................................................................. 150
Table 6‐6. Neutron Yields from Alpha‐Neutron Reactions for Oxides and Fluorides ........................ 151
Table 6‐7. Neutron Yields for Trace Impurities of Uranium............................................................... 152
Table 6‐8. Instrument Response to Uranium Beta Fields .................................................................. 154
Section 12
Table 6‐9. Gamma Flux and Ratios at Various Locations and Sources at Fernald Plant ...................158
Table 6‐10. Performance Test Categories, Radiation Sources, and Test Ranges for the DOELAP and
NVLAP Programs ............................................................................................................... 162
Table 6‐11. Uranium Beta Shielding .................................................................................................... 165
Table 6‐12. Uranium Beta Dose Reduction Factors ............................................................................. 165
Table 9‐1. Summary Table of PAGs, Guidelines, and Planning Guidance for Radiological Incidents .....
(EPA, 2016) ....................................................................................................................... 198
Table 10‐1. Surface Contamination Guidelines ................................................................................... 204
Table of Figures
Figure 2‐1. Uranium Series Decay Chain ................................................................................................. 6
Figure 2‐2. Actinium Series Decay Chain
Figure 6‐3. Changes in Beta Energy Spectra and Shallow Dose Rate from a Natural Uranium Metal
................................................................................................. 7
Figure 4‐1. Protocol for Release of Materials from Radiological Control ............................................. 90
Figure 6‐1. Beta Radiation Readings at Surface of Uranium Metal vs. % Enrichment by Weight ......146
Figure 6‐2. Absorbed Dose Rate as a Function of Depth in Mylar ...................................................... 147
Slab Source Caused by Protective Apparel (Note the bremsstrahlung peak in the low
energy ranges). ................................................................................................................. 148
Figure 6‐4. Meter Readings for a Depleted Uranium Ingot ................................................................ 153
Figure 6‐5. Meter Readings for an Open Drum of UF4 (green salt)..................................................... 153
Figure 6‐6. Measured Angular Response to the INEL TE Survey Meter to Parallel Beams of Beta
Particles from Three Standard Beta Sources. ................................................................... 155
Figure 6‐7. Average Ion Chamber Survey Meter Response by Group to X or Gamma Photon ...............
Radiation ........................................................................................................................... 156
Figure 6‐8. Average GM Survey Meter Photon Energy Response by Group....................................... 157
Figure 6‐9. High Resolution Gamma Spectrum of Slightly Enriched Uranium Oxide (1% 235U) ...............
record with Ge(Li) Detector .............................................................................................. 157
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Section 13
DOE‐STD‐1136‐2017
1 INTRODUCTION
1.1 Purpose and Applicability
This Technical Standard provides operational guidance, practical lessons learned and
experience gained, guides to good practice, and reference information on the safe handling of
uranium. The Technical Standard 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 Part 835). This Technical Standard supplements the DOE G 441.1‐1C,
Radiation Protection Programs Guide for Use with Title 10, Code of Federal Regulations, Part
835 (2008a); DOE Orders; and DOE‐STD‐1098‐2008, Radiological Control (2009c), and its sole
purpose is the protection of workers, the public, and the environment from the hazards that
are inherent in uranium storage, processing, and handling.
This Technical Standard replaces DOE‐STD‐1136‐2009, Guide of Good Practices for
Occupational Radiological Protection in Uranium Facilities (DOE, 2009b), 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 Technical Standard 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 Technical
Standard, in concert with applicable regulatory documents, will help in building a
comprehensive and technically‐defensible radiological control program.
1.2 Definitions
A glossary is provided in Appendix A. In all cases, the definitions provided in this Technical
Standard are consistent with those provided in 10 CFR Part 835, its Guides, and DOE‐STD‐
1098‐2008 (2009c).
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 Dosimetry (Chapter 6), Nuclear Criticality
Safety (Chapter 7), Waste Management (Chapter 8), Emergency Management (Chapter 9), and
Decommissioning (Chapter 10).
1
Section 14
2
DOE‐STD‐1136‐2017
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, to produce radioisotopes, to provide steam for
propulsion, and radiation for research. 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 power 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).
Heavy‐water moderated 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
Section 15
DOE‐STD‐1136‐2017
Table 2‐1. Typical Isotopic Abundances (g of Isotope per 100g of Material)
Isotope Natural
Typical Commercial
Feed Enrichment
Depleted
Specific
Activity (Ci/g)
Neutron
Capture Cross
Section (barns)
238U
235U
234U
99.28
0.72
0.0055
97.01
2.96
0.03
99.8
0.2
0.0007
3.3 E‐7
2.1 E‐6
6.2 E‐3
2.7
101
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 from
the increased 234U than from the 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.
3
DOE‐STD‐1136‐2017
Table 2‐2. Properties of Radionuclides that may be found at Uranium Facilities(a)
Energies (MeV) and Abundances of Major
Radiations
(Low yield radiation is not included)
Nuclide Half‐Life Alpha Beta Gamma
P
ri
m
ar
y
U
ra
n
iu
m
Is
o
to
p
e
s 238U 4.51 x 109 y
4.15 (21%)
4.20 (79%)
235U 7.1 x 108 y
4.21 (6%) 0.144 (11%)
4.37 (17%) 0.163 (5%)
4.40 (55%) 0.186 (57%)
4.60 (5%) 0.205 (5%)
234U 2.47 x 105 y
4.72 (28%)
0.053 (0.12%)
4.77 (72%)
D
e
ca
y
P
ro
d
u
ct
s
234Th 24.1 d
0.103 (21%) 0.013 (9.8%)
0.193 (79%) 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%) 0.026 (2%)
0.287 (12%) 0.084 (10%)
0.288 (37%)
0.305 (35%)
Section 16
Im
p
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ri
ti
e
s
(e
.g
. i
rr
ad
ia
ti
o
n
a
n
d
re
p
ro
ce
ss
in
g
ar
ti
fa
ct
s)
99Tc 2.12x105 y 0.292 (100%)
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 (100%)
232U 72 y
5.26 (31%)
5.32 (69%)
236U 2.34x107 y 4.47 (24%)
(a) From EGG‐2530 (1988).
4
DOE‐STD‐1136‐2017
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. Figures 2‐
1 and 2‐2 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 radiation 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.
5
DOE‐STD‐1136‐2017
Figure 2‐1. Uranium Series Decay Chain
6
DOE‐STD‐1136‐2017
Section 17
Figure 2‐2. Actinium Series Decay Chain
7
DOE‐STD‐1136‐2017
2.1.3 Enrichment
Uranium‐235 enrichment processes selectively increase the 235Uconcentration 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 values are given in Table 2‐3.
Table 2‐3. Uranium Specific Activities
Type Wt. % 235U
Specific Activity of
Mixture (Ci/g)
Natural 0.71 7 x 10‐7
Depleted 0.2 4 x 10‐7
Enriched 2 1 x 10‐6
Highly 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:
;݃ ݅ܥ 10ି ݔ ଶሻܧ 0.0034 ܧൌ ሺ0.4 0.38 ݉ݑ݅݊ܽݎܷ ݄݀݁ܿ݅ݎ݊ܧ ݂ ݕݐ݅ݒ݅ݐܿܣ ݂ܿ݅݅ܿ݁ܵ
݁ ,ݐ݄݃݅݁ݓ ݕܾ ൌ % ଶଷହܷ ܧ ݁ݎ݄݁ݓ 0.72 ݐ݄݊݁݉ܿ݅ݎ݊
⁄
Gaseous diffusion, the predominant existing enrichment technology, causes a greater increase
in 234U concentration than in 235Uconcentration. For example, when 235Ucontent is increased
from 0.72% (natural) to 2.96%, (an increase of approximately a factor of four), 234Ucontent
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 238Uwith 235U,but
more significantly because of the increase in the amount of 234Upresent.
8
Section 18
DOE‐STD‐1136‐2017
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.
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 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.
Example 1
One kilogram of 20% enriched uranium is blended with 1 kilogram of 2% enriched
uranium.
݃/݅ܥ 10ି ݔ ଶሿܧ 0.0034 ܧሻ ൌ ሾ0.4 0.38ܣሺܵ ݕݐ݅ݒ݅ݐܿܣ ݂ܿ݅݅ܿ݁ܵ
݃⁄݅ܥ 10ି ݔ ൌ 9.36 ݃⁄݅ܥ 10ି ݔ ଶ ൌ ሾ0.4 0.38ሺ20ሻ 0.0034ሺ20ሻଶሿܣܵ
݃⁄݅ܥ 10ି ݔ ൌ 1.17 ݃⁄݅ܥ 10ି ݔ ଶ ൌ ሾ0.4 0.38ሺ2ሻ 0.0034ሺ2ሻଶሿܣܵ
The specific activity of the resulting mixture is
9.36 1.17
௫ ൌܣܵ ൨ 10ି ݔ ݃⁄݅ܥ ൌ 5 ݔ 10ି ݃⁄݅ܥ
2
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 238U, 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
9
Section 19
DOE‐STD‐1136‐2017
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.
The Derived Air Concentration (DAC) values for several radionuclides are shown in Table 2‐4.
These include the three uranium isotopes and selected contaminants in recycled uranium.
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 annual limit on
intakes (ALI) 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%.
10
Inhalation DAC, From 10 CFR Part 835, Appendix A
Nuclide Type F µCi/mL (Bq/m3)(a) Type M µCi/mL (Bq/m3) (a)
DOE‐STD‐1136‐2017
Table 2‐4. DACs for Uranium and Selected Contaminants in Recycled Uranium
238U 5 x 10‐10 (2 x 101) 3 x 10‐10 (1 x 101)
235U 5 x 10‐10 (1 x 101) 3 x 10‐10 (1 x 101)
234U 5 x 10‐10 (1 x 101) 2 x 10‐10 (9 x 100)
234Th Not Listed 1 x 10‐7 (3 x 103)
231Th Not Listed 1 x 10‐6 (5 x 104)
99Tc 1 x 10‐6 (5 x 104) 1 x 10‐7 (6 x 103)
237Np Not Listed 8 x 10‐12 (3 x 10‐1)
238Pu Not Listed 6 x 10‐12 (2 x 10‐1)
239Pu Not Listed 5 x 10‐12 (2 x 10‐1)
240Pu Not Listed 5 x 10‐12 (2 x 10‐1)
241Pu Not Listed 2 x 10‐10 (1 x 101)
236U 5 x 10‐10 (1 x 101) 2 x 10‐10 (1 x 101)
(a) See last paragraph of Section 2.5 for discussion of Type F, M, and S.
Example 2
Section 20
One gram of natural uranium contains 239Pu contamination to the extent that the 239Pu
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 (M) = 3 x 10‐10 µCi/mL (use DAC for 238U)
239
Pu derived air concentration (M) = 5 x 10‐12µCi/mL
1 1
ሺܷܥܣܦ െ ሻݐܽܰ
ൌ
ݔ 3 10ିଵ ൌ 3 ݔ 10ଽ
0.001 0.001
൯ݑ൫ଶଷଽܲܥܣܦ
ൌ
10଼ ݔ 10ିଵଶ ൌ 2 ݔ 5
These values represent the relative hazards of the two materials in the mixture.
10଼ݔ2
ൌ ݀ݎܽݖ݄ܽ ݈ܽݐݐ ݂ ݊݅ݐܿܽݎܨ
ሺ210଼ݔሻ ሺ310ݔଽሻ
ൌ 0.06
11
DOE‐STD‐1136‐2017
Therefore, 239Pu at 0.1% of the U‐Nat activity represents 6% of the potential inhalation
dose.
The activity of 1 gram of U‐Nat = 2.5 x 104dps
Therefore, the 239Pu activity in the 1 gram of U‐Nat.= 0.001 x 2.5 x 104 = 25 dps
239
The specific activity of Pu is 2.27 dps/nanogram:
ݔ
ݑܲ ݉ܽݎ݃݊ܽ݊ 1
⁄25 ݀ ܷ ⁄݃ ݏ
݉ܽݎ݃݊ܽ݊/ݏ݀ 1
ൌ 11 ݊ ܽ ܷ ݃ ݑܲ ݏ݉ܽݎ݃݊
Therefore, 0.1% 239Pu activity fraction corresponds to 11 parts per billion on a mass
basis.
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
Section 21
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.
12
DOE‐STD‐1136‐2017
The isotope in recycled uranium presenting the greatest potential radiological hazard from
external sources is 232U. 232U is 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
Section 22
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 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 232Umay 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
13
DOE‐STD‐1136‐2017
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, heavy neoprene gloves, and safety glasses). 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
Section 23
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 doses to the skin from
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 health 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 alloy‐clad ceramic uranium dioxide and uranium carbide fuels were found to have
acceptable characteristics and are in common use.
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.
14
DOE‐STD‐1136‐2017
Section 24
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 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 pellets 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
15
Section 25
2.3
DOE‐STD‐1136‐2017
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 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
chips and turnings are pyrophoric and tend to catch fire.
Uranium metal may be dissolved using nitric acid, which is also used to passivate ("pickle") the
metal to inhibit oxidation.
Radiological Characteristics and Effects
Uranium isotopes decay by alpha particle emission and some also emit low‐energy gamma
rays. For Types M and S material (See last paragraph of Section 2.5 for discussion of Type F, M
and S), 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 and lens of the eye dose resulting from external radiation exposures, due mainly to the
2.29 MeV (Emax) beta from 234mPa.The surface dose rates shown in Table 2‐5 result primarily
from beta radiation from decay products. The dose rates decrease quickly with distance
because of the attenuation of the beta radiation and the small yield of the gamma radiation.
Table 2‐5. Beta Surface Dose Rates from Equilibrium Thickness of Uranium Metal and
Compounds(a)
Source
Beta Surface Dose Rate
(mrad/hr)
U‐Nat metal slab(b)
UO2
UF4
UO2(NO3)26H20
UO3
U3O8
UO2F2
Na2U2O7
233
207
179
111
204
203
176
167
(a) Beta surface dose rate in air through a
polystyrene filter 7 mg/cm2 thick.
(b) freshly polished, no oxide
Because some uranium decay products have short half‐lives (on the order of days), those
16
DOE‐STD‐1136‐2017
decay products will usually be present with uranium during processing. An assumption of
Section 26
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 beta levels from decay products may be many orders of magnitude greater than the
exposure rates from the uranium.
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 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.
17
Section 27
DOE‐STD‐1136‐2017
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 International Commission on Radiological Protection (ICRP)
Publication 66 (1994a) 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.
2.3.2.2 Ingestion
Appropriate uranium contamination controls should prevent ingestion of uranium.
Nevertheless, the potential exists for accidental ingestion of uranium. Particles removed from
Section 28
the respiratory tract by ciliary action are transferred to the gastrointestinal tract. Particles
18
2.4
DOE‐STD‐1136‐2017
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 (1966). This model calculates
material transferred from the GI tract to the blood based on solubility classes (ICRP, 1979;
IAEA, 1994) or based on a single value for all compounds, as described in ICRP Publication 69
(1995a).
Distribution of uranium transferred into the bloodstream is calculated using a once‐through
metabolic model. ICRP Publication 30 (1979) also provides values for this distribution and
excretion to calculate committed doses and long‐term tissue retention. Recent models
(Wrenn et al., 1994; ICRP, 1995a) have been developed to include recycling of uranium back
into the blood.
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. Epidemiological and toxicological animal studies of
oral exposure to relatively soluble naturally occurring isotopic mixtures appear demonstrate
that the health effects are chemical toxicity to tissues; while those from inhalation exposure
may include a slight radiological component, especially if the exposure is chronic (ATSDR,
2013). For inhalation of relatively insoluble uranium compounds, the radiological hazard can
be the most limiting due to an increased residence time in the lungs and low fractional
absorption to blood (Leggett, et. al., 2012).
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 (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
Section 29
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.
Data on human exposures and the effect of various intakes of uranium are summarized in
Table 5‐7. These data indicate that a single intake of 8 mg of natural uranium would be well
19
‐
DOE‐STD‐1136‐2017
below the level that could cause permanent kidney damage in most individuals, and that 4 mg
intake would likely cause no observable effects. The urine levels for situations in which
chemical toxicity might be of concern are based on interpretation of the data (McGuire, 1991).
Table 2‐6. Health Effects from Acute Intake of Soluble Uranium(a)
HEALTH EFFECTS
URANIUM per kg
BODY WT
(mg U kg 1)(b)
URANIUM IN 70 kg
PERSON
(mg)
URANIUM INTAKE BY 70
kg PERSON
(mg)(c)
50% Lethality
Threshold for permanent
renal damage
Threshold for transient
renal damage
No effect
1.63
0.3(d)
0.058
0.03
114
21
4.06
2.1
230
40
8
4
(a) (ANSI N13.22‐2013); based on review (McGuire, 1991).
(b) Based on review (Just and Emler, 1984), except where noted.
(c) For this table, intake is defined as the total amount of material inhaled into the body. It includes material
immediately exhaled in addition to material absorbed within the body. For small uranium particles in
soluble form, about half of the intake will be absorbed by the body according to ICRP Publication 68
(1994b).
(d) See discussion in (Just and Elmer, 1984)
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 DOE regulated facilities, the more
conservative of the two standards (OSHA or ACGIH) must be used unless enrichment and
solubility dictate more stringent controls based on radiological concerns. Table 2‐6 lists
airborne concentration limits for transportable uranium that have been published by various
organizations.
20
Section 30
DOE‐STD‐1136‐2017
Table 2‐7. 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, 1992b)
ACGIH(a) 0.2 Threshold Limit Values and
Biological Exposure Indices for
2005, American Conference of
Governmental Industrial
Hygienists (ACGIH, 2005)
OSHA(b) 0.05 (soluble) 29 CFR § 1910.1000
0.25 (insoluble)
NIOSH 0.05 National Institute for
Occupational Safety and Health
(NIOSH)
(a) ACGIH also has a short term exposure limit of 0.6 mg/m3. This is based on a 15
minute time weighted average exposure.
(b) Preferred/recommended limit.
Past limits for single acute inhalation intakes have been set by the ICRP in its Publication 6,
(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 (1984)
derived acute inhalation intake limits of 15 and 80 mg for Type F and Type M 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 Part 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 68 (1994b) metabolic model for Type F 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 Part 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
a. Lack of data on the effects of combined exposures to UO2F2 and HF.
b. Lack of detailed information on effects of short‐term exposures to soluble/transportable
21
DOE‐STD‐1136‐2017
uranium in the range from 100‐1000 mg/m3.
c. 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
Section 31
50). Researcher consensus resulted in the kidney burdens (in µg U/g Kidney) listed in Table 2‐
7.
Table 2‐8. Uranium Levels for Various Effects
Effect
Kidney Burden,
µg U/g Kidney
Total Kidney Burden,
mg U
Intake,
mg
No effect 1.1 0.337 6.5
Maximal Nonlethal 2.2 0.71 13
LD50 54.8 16.79 322
The kidney burden values can be used to derive an intake based on the ICRP Publication 78
(1997), model for uranium metabolism (62% of inhaled Type F uranium is taken up into the
bloodstream (82% deposited – 62% systemic and 20% to GI tract) and 8.5% of that goes to the
kidneys). For example, the "no effect" value in Table 2.7 corresponds to an intake of (1.1μg
U/g)(310 g/kidney)(1000 μg/mg)/((0.62 fraction systemic)(.085 fraction systemic going to
kidney) = 6.5 mg.
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.
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 (0.62)
fk = bloodstream fraction entering kidneys (0.085) and
22
http:�g/mg)/((0.62
DOE‐STD‐1136‐2017
R = Kb where
where
R = kidney removal rate (mg/day)
= 0.099 (day‐1) (ICRP, 1997)
Kb = amount in the kidney (mg)
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 0.099 = 0.033 mg/day
K = Br (m3/day) x Ca(mg/m3) x (0.62) x (0.085)
K= R = 0.033 mg/day
Br x Ca x (0.62) x (0.085) = 0.033 mg/day
Br x Ca = 0.63 mg/day
Standard man breathes 9.6 m3of air in an 8‐hour day, so the resulting concentration limit is
0.63/9.6 = 0.066 mg/m3. This is 30% 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., Type F) 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
Section 32
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.
23
DOE‐STD‐1136‐2017
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 F) 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
Section 33
with potentially life‐threatening consequences and would indicate the need for immediate
hospitalization.
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, 1992a).
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. ICRP Publication 88 (2001) gives dose coefficients for the embryo,
fetus and newborn of females following intake of selected radionuclides. It also provides a
review of biokinetic and dosimetric models for calculating doses to the offspring of mothers
following intakes by the mother before or during pregnancy.
2.5 Chemical versus Radiological Hazards
Both the chemical and radiological hazards of uranium are moderate compared to those of
24
DOE‐STD‐1136‐2017
other industrial materials and radionuclides. Table 2‐4 provides 10 CFR Part 835 derived air
concentration values for selected radionuclides. Table 2‐8 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.
Table 2‐9. 2005 ACGIH Threshold Limit Values (TLVs) for Selected Metals
TLV
Metal TLV‐TWA, mg/m3 TLV‐STEL, mg/m3
Uranium 0.2 0.6
Beryllium 0.002 0.01
Lead 0.05 ‐‐
Mercury, elemental and
inorganic forms
0.025 ‐‐
Arsenic 0.01 ‐‐
TLV‐TWA = Threshold Limit Value, Time‐Weighted Average
TLV‐STEL = Threshold Limit Value, Short‐term Exposure Limit
Section 34
The relative activities of the primary uranium isotopes are also significantly affected by the
degree of enrichment. The 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 60 (1991a) classifies all materials into three material types ‐ F, M, and S. Type F is
most transportable (pulmonary removal half‐time of days), Type S the least transportable
(removal half‐time of years), and Type M 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 2‐9 lists several
common uranium compounds and their assigned material types.
25
DOE‐STD‐1136‐2017
Table 2‐10. Inhalation Classification for Some Uranium Compounds
Uranium Compound Chemical Name Material Type
Uranium hexafluoride UF6 Type “F”
Uranyl fluoride UO2F2 Type “F”
Uranyl nitrate UO2(NO3)2 Type “F”
Uranyl acetate UO2(C2H3O2)2 Type “F”
Uranyl chloride UO2Cl2 Type “F”
Uranyl sulfate UO2SO4 Type “F”
Uranium trioxide UO3 Type “M”
Uranium tetrafluoride UF4 Type “M”
Uranium oxide U3O8 Type “S”(b)
Uranium dioxide UO2 Type “S”(b)
Ammonium diuranate (NH4)2 + U2O7 Type “M” (a)
Uranium aluminide UAlx Type “S”
Uranium carbide UC2 Type “S”
Uranium‐zirconium alloy UZr Type “S”
High‐fired uranium dioxide UO2 Type “S”(b)
(a) Ammonium diuranate is known to contain uranium as UO3, and should not
be assigned to a single inhalation class.
(b) The solubility of uranium oxides is very dependent on heat treatment. The
rate of oxidation may also affect the solubility. It is recommended that
solubility studies be performed to characterize the actual materials present.
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 chemical 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
Section 35
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 receive
100 millirem committed effective dose (CED) from all internal and external occupational doses
received in a years. Therefore, it is prudent to calculate organ doses and CED 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
26
DOE‐STD‐1136‐2017
comprehensive dosimetry/control program can prove invaluable in 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). For comparison, the
more conservative of the two chemical toxicity standards (OSHA or ACGIH) is used to compare
the chemical hazard with the radiological hazard. The ICRP Publication 68 (1994b) dose
coefficients and ICRP 78 (1997) models are used in the following examples to determine the
relative hazards for acute and chronic exposure situations. Table 2‐10 summarizes the dose
coefficients used in the following examples.
Table 2‐11. Dose Coefficients for Determining Enrichments above which Radiological
Hazards Become Limiting
Effective
Dose
Coeff,
Sv/Bq
Effective
Dose
Coeff,
Sv/Bq
Effective
Dose
Coeff,
Sv/Bq
Organ
Dose
Coeff,
Sv/Bq
Organ
Organ
Dose
Coeff,
Sv/Bq
Organ
Organ
Dose
Coeff,
Sv/Bq
Organ
F M S F M S
U‐234
U‐235
U‐238
6.40E‐07
6.00E‐07
5.80E‐07
2.10E‐06
1.80E‐06
1.60E‐06
6.80E‐06
6.10E‐06
5.70E‐06
1.10E‐05
1.10E‐05
1.00E‐05
BS
BS
BS
1.60E‐05
1.40E‐05
1.30E‐05
LG
LG
LG
7.50E‐05
6.90E‐05
6.50E‐05
ET
ET
ET
ET – Extrathoracic airways LG – Lung BS – Bone Surface
To determine which hazard is limiting for a chronic exposure, the chemical toxicity air
concentration limit (0.2 mg/m3 for material type “S” and 0.05 mg/m3 for material types “F”
and “M”) and a calculated derived air concentration (DAC), based on percent enrichment, are
used. For material type “F” the DAC is based on 50 rems to the bone surface, for material
type “M” the DAC is based on 5 rems effective dose, and for material type “S” the DAC is
based on 50 rems to extrathoracic portion of the respiratory tract. The dose coefficients from
Table 2‐10 are used to calculate a DAC for varying enrichments. The enrichment at which the
DAC is equal to the chemical toxicity limit forms the "dividing line" between chemical and
Section 36
radiological limits as the limiting hazard. Generally, exposures to higher enrichments are
limited by radiological limits; exposures to lower enrichments by chemical limits. Example 3a
provides the methodology for determining the "dividing line" enrichment for the chronic
exposure scenario. The same calculation is done using the 100 millirem monitoring threshold.
For an acute exposure scenario, the amount of an intake which would result in exceeding the
fifteen minute time weighted chemical toxicity short term exposure limit for material types
“F”, “M”, and “S” uranium (i.e., 0.6 mg/m3 x 0.3 m3 air intake per fifteen minute ) is
27
DOE‐STD‐1136‐2017
determined. This amount is compared to the amount of an intake which would result in
exceeding the radiological limit. Again, for material type “F” the radiological limit is based on
50 rems to the bone surface, for material type “M” the radiological limit is based on 5 rems to
the whole body, and for material type “S” the radiological limit is based on 50 rems to
extrathoracic portion of the respiratory tract. Example 3b provides the methodology used for
an acute exposure scenario. Table 2‐11 summarizes the “dividing line” enrichments, above
which radiological limits are controlling. The same calculation is done using the 100 millirem
monitoring threshold.
Example 3a ‐ General Solution, Chronic Exposure
Step 1. For increasing 235U enrichments, and consequential 234U enrichment, determine the
activity fraction (AF) from 234U, 235U, and 238U in the enriched uranium.
∗ ଶଷସݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷସ ൌܨܣଶଷସܣܵ
∗ ଶଷସݐ݄݊݁݉ܿ݅ݎ݊ܧ ∗ ଶଷହݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷସ ܣܵ ∗ ଶଷ଼ݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷହ ܣܵ ଶଷ଼ܣܵ
∗ ଶଷହݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷହ ൌܨܣଶଷହܣܵ
∗ ଶଷସݐ݄݊݁݉ܿ݅ݎ݊ܧ ∗ ଶଷହݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷସ ܣܵ ∗ ଶଷ଼ݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷହ ܣܵ ଶଷ଼ܣܵ
∗ ଶଷ଼ݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷ଼ ൌܨܣଶଷ଼ܣܵ
∗ ଶଷସݐ݄݊݁݉ܿ݅ݎ݊ܧ ∗ ଶଷହݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷସ ܣܵ ∗ ଶଷ଼ݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷହ ܣܵ ଶଷ଼ܣܵ
Where:
Enrichment is the percent by weight.
SA234 = specific activity of 234U = 2.30E+08 Bq/g
SA235 = specific activity of 235U = 79312 Bq/g
SA238 = specific activity of 238U = 12329 Bq/g
మయర
≅ ଶଷସݐ݄݊݁݉ܿ݅ݎ݊ܧ
൫.ହହ ௧௨ ௧
మయఱ
൯
∗ ଶଷହܷ ݁݊1.2 ∗ ݐ݄݊݁݉ܿ݅ݎ(a)
ሺ.ଶ ௧௨ ௧
(a) Gaseous diffusion enrichment of 235U results a proportionally greater enrichment of
234U. The 1.2 factor takes into account the gaseous diffusion process causing a greater
increase in 234U than in 235U (See Section 2.1.3).
ଶଷସ െݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷ଼ ൌ 100 െݐ݄݊݁݉ܿ݅ݎ݊ܧ ଶଷହݐ݄݊݁݉ܿ݅ݎ݊ܧ
Step 2. Using the dose coefficients from Table 2‐10 and the activity fractions, for increasing
enrichments, determine activity enriched uranium organ and effective dose coefficients for
material types “F”, “M” and “S”.
݂݂݁ܥ ݁ݏܦ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
ൌ ∗ ଶଷସܨܣ ݁ݏ݀ ܿ ݂݁ ଶ݂ଷସ ∗ ଶଷହܨܣ ݁ݏ݀ ܿ ݂݁ ଶ݂ଷହ ଶଷ଼ܨܣ
∗ ݁ݏ݀ ܿ ݂݁ ଶ݂ଷ଼
For example, for 5% enriched uranium (i.e., 5%, by weight, of the uranium is235U), material
type “S” dose to the extrathoracic portion of the respiratory tract:
Section 37
28
DOE‐STD‐1136‐2017
Using the equations from Step 1, it is determined that, for 5% enriched uranium,
87% of the activity is from 234U, 3% from 235U, and 10% from 238U.
From this we can determine a dose coefficient for 5% enriched uranium using the
above equation:
⁄ ܵ െ 5 ܧൌ 0.87 ∗ 7.5 ݂݂݁ܥ ݁ݏܦ ݐ݄݊݁݉ܿ݅ݎ݊ܧ 5%⁄ ݍܤ ݒ 0.03 ∗ ܧ6.9 െ 05 ܵ ݒ ݍܤ
⁄0.1 ∗ ܧ6.5 െ ൌ ⁄ݍܤ ݒ5ܵ ܧ7.38 െ ݍܤ ݒܵ 5
Step 3. The ALI is determined, for different enrichments using the % enrichment dose
coefficient determined in Step 2. For material type “F” the % enrichment ALI is based on 50
rems to the bone surface, for material type “M” it is based on 5 rems effective dose, and for
material type “S” on 50 rems to extrathoracic portion of the respiratory tract.
ݐ݅݉݅ܮ ݁ݏܦ
ൌ ܫܮܣ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
݂݂݁ܿ ݁ݏ݀ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
The % enrichment ALI, which is units of Bq, is converted to a mass using the SA %
enrichment formula and is divided by the volume of air a worker breathes in a work year
(2000 hours), 2400 m3, to give a mass airborne concentration value.
∗ ሻଶሿݐ݄݊݁݉ܿ݅ݎሻ 0.0034ሺ݁݊ݐ݄݊݁݉ܿ݅ݎൌ ሾሺ0.4 0.38ሺ݁݊ ݐ݄݊݁݉ܿ݅ݎ݊ܧ % ܣܵ ܧ1.0 െ 6;
Where enrichment is the fraction by weight of 235U, expressed as a percentage.
Note that the value of SA_% Enriched as calculated by the formula is in units of µCi/g
(Equation from 10 CFR Part 20 Appendix B, footnote 3)
ሺ% ܫܮܣ ݐ݄݊݁݉ܿ݅ݎ݊ܧሻ
ሺ݃ ݉ଷሻ ൌ ݊݅ݐܽݎݐ݊݁ܿ݊ܥ ݎ݅ܣ⁄
݁݇ܽݐ݊ܫ ݁݉ݑ݈ܸ ݎ݅ܣ ∗ ݐ݄݊݁݉ܿ݅ݎ݊ܧ% ܣܵ
The mass airborne concentration value is used for comparison with the chemical toxicity
limit.
Continuing the above example: the ALI for 5% enriched uranium, material type “S” dose to
the extrathoracic portion of the respiratory tract:
ݒܵ 0.5 1 μ݅ܥ
∗ ൌ ܫܮܣ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
ݍܤ 37000
ൌ 0.18 μ݅ܥ
⁄
݃/݅ܥെ 6 μ ܧെ 6 ൌ 2.38 ܧൌ ሺ0.4 0.38 ∗ 5 0.0034 ∗ 5ଶሻ ∗ 1.0 ݐ݄݊݁݉ܿ݅ݎ݊ܧ %5 ܣܵ
ܧ7.38 െ ݍܤ ݒ5ܵ
0.18 μ݅ܥ
ൌ ݊݅ݐܽݎݐ݊݁ܿ݊ܥ ݎ݅ܣ⁄
⁄ ∗ 2400 ݉ଷ ∗
1000 ݉݃
ൌ 0.032 ݉݃ ݉ଷ
ܧ2.38 െ 6μ݅ܥ ݃ 1 ݃
29
DOE‐STD‐1136‐2017
Step 4. This value is compared to the chronic chemical toxicity air concentration limit (0.2
mg/m3 (ACGIH TLV) for material type “S” and 0.05 mg/m3 (OSHA soluble PEL) for material
types “F” and “M”. As the enrichment increases the air concentration values determined
for radiological control decrease. The enrichment where they fall below the chemical
toxicity limits becomes the “dividing line” where radiological limits become more restrictive.
Section 38
In the above example the air concentration value, 0.032 mg/m3, is lower than the chemical
toxicity value and is therefore limiting (for type “S” material, chronic exposure, radiological
considerations are limiting for all enrichments).
Step 5. The above process is repeated using the 100 millirem monitoring threshold in lieu of
the annual dose limit. For these calculations the effective dose coefficients are used for all
isotopes.
Example 3b ‐ General Solution, Acute Exposure
Steps 1 and 2: Same as in example 3a.
Step 3. This is similar to Step 3 in example 3a, with the exception that % enrichment ALI is
not divided by the volume of air a worker breathes in a work year, 2400 m3.
The annual limit on intake (ALI) is determined, for different enrichments using the %
enrichment dose coefficient determined in Step 2. For material type “F” the % enrichment
ALI is based on 50 rems to the bone surface, for material type “M”, it is based on 5 rems to
the whole body, and for material type “S”, on 50 rems to extrathoracic portion of the
respiratory tract. The resultant value, in Bq, is converted to a mass using the SA % enriched
formula. The result is a mass intake limit.
ݐ݅݉݅ܮ ݁ݏܦ
ൌ ܫܮܣ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
݂݂݁ܿ ݁ݏ݀ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
ൌ ݐ݅݉݅ܮ ݁݇ܽݐ݊ܫ ݏݏܽܯ
ܫܮܣ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
ݐ݄݊݁݉ܿ݅ݎ݊ܧ % ܣܵ
From example 3a: the ALI for 5% enriched uranium, material type “S” dose to the
extrathoracic portion of the respiratory tract:
ݒܵ 0.5 1 μ݅ܥ
∗ ൌ ܫܮܣ ݐ݄݊݁݉ܿ݅ݎ݊ܧ %
ݍܤ 37000
ൌ 0.18 μ݅ܥ
ܧ7.38⁄ െ ݍܤ ݒ5ܵ
0.18 μ݅ܥ
∗
1000 ݉݃
ൌ ൌ ݐ݅݉݅ܮ ݁݇ܽݐ݊ܫ ݏݏܽܯ 75.8 ݉ ݃
ܧ2.38⁄ െ 6 μ݅ܥ ݃ 1 ݃
Step 4. This value is compared to the product of the volume breathed in fifteen minutes
times the short term exposure limit for chemical toxicity (0.6 mg/m3 (OSHA STEL) * 0.3 m3
30
DOE‐STD‐1136‐2017
breathed per fifteen minutes = 0.16 mg for material types “F”, “M”, and “S”. As the
enrichment increases the mass acute intake limit determined for radiological control
decreases. The enrichment where they fall below the chemical toxicity limits becomes the
“dividing line” where, for an acute exposure, radiological limits become more restrictive.
In the above example the mass acute intake limit, 75.8 mg, is higher than the chemical
toxicity value and therefore the chemical toxicity limit is controlling for this enrichment (for
acute exposures, compared to the radiological limit, chemical toxicity considerations are
always limiting).
Step 5. The above process is repeated using the 100 millirem monitoring threshold in lieu of
the annual dose limit. For these calculations the effective dose coefficients are used for all
isotopes.
Section 39
Table 2‐12. Impact of Monitoring at 100 Millirem Enrichments above which Radiological
Limits Predominate (Calculations Not Shown)
Acute Chronic
Material
Type
Using 100% of
Radiological
Limit
Using 2% of
Radiological
Limit
Using 100% of
Radiological
Limit
Using 2% of
Radiological Limit
F
M
S
(a)
(a)
(a)
(a)
99.6%
41.8%
21.8%
12.0%
(b)
(b)
(b)
(b)
(a) Chemical toxicity limits are limiting at all enrichments.
(b) Radiological limits are limiting at all enrichments.
As shown in Table 2‐11, for chronic exposures, the 100 millirem monitoring threshold (i.e., 2%
of radiological limit) results in radiological conditions, requiring evaluation of the need to
monitor radiological intakes, as a controlling factor. For chronic exposures the radiological
dose limits are controlling for type “S” material for all enrichments, and for lower enrichments
for types “M” and “F” material. For acute exposure situations, the radiological monitoring
threshold is a controlling factor for higher enrichments for types “M” and “S” material. For all
other acute exposure situations, the chemical toxicity limits, which are based on a fifteen
minute or less exposure, are more controlling than the radiological limits, which are based on
an annual exposure.
Natural Uranium Balance in Man
Uranium is present in trace quantities throughout the environment. As a result, man ingests
about 2 µg of natural uranium each day in food and fluids. A similar quantity is excreted each
day in the feces and urine. The uranium balance for reference man is presented in Table 5‐8.
31
2.6
DOE‐STD‐1136‐2017
Table 2‐13. Uranium Balance for Reference Man(a)
Intake:
Food and fluids:
Inhalation:
1.9
7.0 E‐3
µg/day
µg/day
Losses:
Feces:
Urine:
Other (hair)
1.4 ‐ 1.8
0.05 ‐ 0.5
0.02
µg/day
µg/day
µg/day
(a) ICRP Publication 23 (1975)
The range of intake and losses has been observed to vary over several orders of magnitude,
depending upon the uranium concentration in foods and in the water supply.
2.7 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, e.g., back injuries from lifting, dropping heavy parts on feet. 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.
2.7.1 Hydrogen Fluoride
Hydrogen fluoride is an extremely corrosive acid that is relatively volatile in its anhydrous
Section 40
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.
32
DOE‐STD‐1136‐2017
2.7.2 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 3 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).
2.7.3 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
Section 41
reactant should include design features (e.g., H2 monitors, roof vents) 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 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).
2.7.4 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
33
DOE‐STD‐1136‐2017
is still burning, it will resume burning visibly. DOE‐HDBK‐1081‐2014, Primer on Spontaneous
Heating and Pyrophoricity (2014b), contains additional guidance.
34
3
DOE‐STD‐1136‐2017
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 Part 835. DOE has
Section 42
provided supporting and clarifying guidance in the DOE G 441.1‐1C, DOE‐STD‐1098‐2008
(2009c), and DOE Radiological Control Technical Positions. Other related source documents
include publications of the EPA, NRC, 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 441.1‐1C (2008a) suggests the following,
based on the content of 10 CFR Part 835:
a. Organization and Administration
b. ALARA Program
c. External Dosimetry Program
d. Internal Dosimetry Program
e. Area Monitoring and Control
f. Radiological Controls
g. Emergency Exposure Situations
h. Nuclear Accident Dosimetry
i. Records
j. Reports to Individuals
k. Radiation Safety Training
l. Limits for the Embryo/Fetus
Each of these functional elements is discussed in more detail below.
35
DOE‐STD‐1136‐2017
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 G441.1‐1C (2008a) and DOE‐STD‐1098‐2008 (2009c) provide detailed
guidance on implementing these requirements.
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 Part 835 and shall be approved by DOE (10 CFR § 835.101). Any
changes that decrease the effectiveness of the RPP shall be approved by DOE before
implementation (10 CFR § 835.101).
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 than every 36 months (10 CFR § 835.102). An effective quality assurance
Section 43
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 activities 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 Part 830, Nuclear Safety Management
(2011f), DOE O 414.1D, Quality Assurance (2013e), and their associated guides. Specific
guidance applicable to RPPs is provided in DOE G 441.1‐1C (2008a).
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, which 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 formal plans
36
DOE‐STD‐1136‐2017
and measures for applying the as low as reasonably achievable (ALARA) process to
occupational exposure (10 CFR § 835.101(c)).
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 Part 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 Part 835. Written procedures should be
developed and employed under the following circumstances:
Section 44
a. When worker health and safety are directly affected
b. When the expected outcome for the process or operation requires that a specific
method be followed
c. When the process or operation is infrequently used and competence training cannot
assure adequate implementation
d. 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. DOE G 441.1‐1C provides 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 N13.6‐2010, Practice for Occupational
Radiation Exposure Records Systems (2010) 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
37
DOE‐STD‐1136‐2017
ensure that the required reviews and revisions occur.
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 DOE‐STD‐1098‐2008 (2009c)). 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
Section 45
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.
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.
38
DOE‐STD‐1136‐2017
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 personnel in the performance 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
Section 46
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 qualified staff of RCTs should be present. All RCT training
should be accomplished in accordance with the DOE‐STD‐1098‐2008 (2009c) and DOE‐HDBK‐
1122‐2009, Radiological Control Technician Training Program (2011c).
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 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 DOE‐STD‐
39
DOE‐STD‐1136‐2017
1098‐2008 (2009c) 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 426.2,
Personnel Selection, Training, Qualification, and Certification Requirements for DOE Nuclear
Facilities (2013b). DOE‐STD‐1098‐2008 (2009c) provides additional guidance. Each uranium
Section 47
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
a. DOE‐HDBK‐1130‐2008, Radiological Worker (RW) Training
b. DOE‐HDBK‐1131‐2007, General Employee Radiological Training
c. DOE‐STD‐1107‐97, CN1, Knowledge Skills and Abilities for Key Radiation Protection
Positions at DOE Facilities
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.
40
DOE‐STD‐1136‐2017
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
Section 48
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 Part 835 and DOE O 458.1 establish the policy of
maintaining doses ALARA for workers and the public, respectively, resulting from radiation
from DOE operations. 10 CFR § 835.101(c) requires that plans and programs implementing the
ALARA process be prepared, and 10 CFR § 835.704(b) requires that records must be must be
maintained to demonstrate the implementation of ALARA. DOE G 441.1‐1C (2008a) and DOE‐
STD‐1098‐2008 (2009c) 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.
41
DOE‐STD‐1136‐2017
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.
3.2.2.3 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
Section 49
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.
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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 HEPA exhaust fans, ductwork and filters,
Section 50
and special fixtures to hold highly radioactive materials requiring detailed inspection, repair,
modification, or fabrication. Such devices can reduce radiation exposures by controlling
contamination, 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.
3.2.2.6 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
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radiological work has occurred. Such trends should be examined at least quarterly to permit
Section 51
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 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 441.1‐1C (2008a).
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Section 52
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 441.1‐1C (2008a).
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 441.1‐1C (2008a).
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 441.1‐1C (2008a).
3.2.5.1 Radiological Surveys and Data Trending
Sections 835.401 ‐ 835.403 of 10 CFR Part 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.
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
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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 radiation levels, beams, or hot
spots
d. Ensure radiological conditions are acceptable and documented prior to, during, and at
Section 53
the completion of work that may cause changes in radiation levels to occur
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 as follows:
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
f. When requested by the personnel performing the activity
A sufficient number of points should be surveyed to adequately assess the radiological status
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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, the
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)
d. Dose rates
e. Estimated doses to surveyors (from direct‐reading dosimeters, if applicable)
f. Date and time the survey was performed
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 10 CFR §
835.701(b) which requires retention of records until final disposition is authorized by DOE.
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.
Section 54
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
d. Identify the origin of radiation exposures in the plant by location, system, or component
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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
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.
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 N317‐1980,
Performance Criteria for Instrumentation Used for In‐Plant Plutonium Monitoring (1980).
Performance specifications are also given in IEEE/ANSI N323AB‐2013 (2014a), Radiation
Section 55
Protection Instrumentation Test and Calibration, Portable Survey Instruments; IEEE/ANSI
N42.17A‐2003, Performance Specifications for Health Physics Instrumentation ‐ Portable
Instrumentation for Use in Normal Environmental Conditions (2004a); IEEE/ANSI N42.17C‐
1989, Performance Specifications for Health Physics Instrumentation ‐ Portable
Instrumentation for Use in Extreme Environmental Conditions for portable radiological control
instrumentation (1990b); and IEC Publication 60325:2002, Radiation Protection
Instrumentation – Alpha, beta and alpha/beta (beta energy >60 keV) contamination meters
and monitors (2002b), for alpha and beta contamination meters and monitors. Criteria for air
monitoring instrumentation are provided in ANSI/HPS N13.1‐2011, Sampling and Monitoring
Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities
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(2011a); IEC Publication 60761, Equipment for Continuously Monitoring Radioactivity in
Gaseous Effluents (2002a); and ANSI N42.17B‐1989, Performance Specifications for Health
Physics Instrumentation ‐ Occupational Airborne Radioactivity Monitoring Instrumentation
(1990a). Criticality alarm systems are discussed in ANSI/ANS 8.3‐1997, Criticality Accident
Alarm System (1997). The criteria discussed in the following sections are specified in these
standards as referenced.
Portable Monitoring Instruments
ANSI N317 (1980) 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.
d. The instrument should have a minimum battery lifetime of 200 hours of continuous
operation. ANSI N42.17A (2004a) 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 (1980) states the minimum detection capability for alpha monitoring instruments
Section 56
ideally should be 220 dpm/100 cm2 of surface area and should not be more than 500 dpm/100
cm2. This criteria 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 criteria stated in ANSI N317 (1980)
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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 (2004a) has a broader scope than ANSI N317 (1980), 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.
Section 57
For direct alpha contamination surveys, the use of audible signals (headphones or speaker)
greatly facilitates the detection of "hot spots." IEC Publication 60325 (2002b) provides
additional guidance on the uniformity of probe response for alpha and beta contamination
meters. Surface sensitivity measurements are also discussed in this standard.
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 (1980). 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
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