DOE-STD-1121-2008 (Reaffirmed 2022), Internal Dosimetry

DOE-STD-1121-2008 Change Notice No.1 Reaffirmation November 2022 DOE STANDARD INTERNAL DOSIMETRY U.S. Department of Energy AREA SAFT Washington, DC 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. NOT MEASUREMENT SENSITIVE DOE-STD-1121-2008 ii This document is available on the Department of Energy Technical Standards Program Web Site at https://www.standards.doe.gov/ DOE-STD-1121-2008 iii Table of Changes Page/Section Original Change Throughout Correct Grammatical Errors Throughout HS-11 EHSS-11 Throughout Title 10, Code of Federal Regulations, Part 835 (DOE 2007a) Title 10, Code of Federal Regulations, Part 835 (DOE 2017) Throughout “RadCon Standard;” DOE 2008a “RadCon Standard;” DOE-STD- 1098-2017 Throughout rems rem Throughout Implementation Guide for Use with 10 CFR 835 (DOE 2008b) Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a) Throughout ANSI N323a-1997 ANSI N323ab-2013 Throughout ANSI Z88.2-1992 ANSI Z88.2-2015 Throughout HPS N13.1-2011 ANSI/HPS N13.1-2011 Throughout HPS N13.6-2010 ANSI/HPS N13.6-2010 Throughout HPS N13.12-1999 ANSI/HPS N13.12-2013 Throughout HPS N13.30-1996 ANSI/HPS N13.30-2011 Throughout HPS N13.39-2001 ANSI/HPS N13.39-2011 Throughout ANSI N42.17A-1994 ANSI N42.17A-2003 Throughout NRC (1993b) NRC (2015) Throughout ICRP-32 (1981) ICRP-65 (1993a) Throughout DOE/EH-01737 DOE-HDBK-1216-2015 Section 1.0, p. 1 Added: “roentgen equivalent man” and “sievert” before using the acronym Section 1.1, p. 1 Internal Dosimetry in the Implementation Guide for Use with 10 CFR 835 (10 CFR 835 Implementation Guide) (DOE 2008b) Internal Dosimetry Program, in the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE G 441.1-1C Chg 1) (DOE 2011a) Section 1.4, p. 2 100,000 radiation workers 80,000 radiation workers Section 2.0 Nonsubstantive updates to definitions to reflect current consensus standards Section 2.0, p. 5 Internal Dosimetry Internal Dosimetry Program Section 2.2, p.11, third paragraph Lead-212 212Pb from 212Bi and 212Po. DOE-STD-1121-2008 iv Table of Changes (cont’d) Section 3.1, p. 20 The technical basis documentation should be reviewed periodically and updated as necessary to ensure that the scientific bases are appropriate for current conditions. Section 3.1.3, p. 22, first paragraph Generally, it is acceptable for the internal dosimetry program for an individual to be discontinued Generally, it is acceptable for an individual to be no longer monitored as part of the internal dosimetry program Section 3.1.5, p. 23, tenth bullet individual-specific and facility- specific factors that are expected the basis for determining which individual-specific and facility- specific factors are expected Section 3.1.6, p. 24, first bullet Delete: (RadCon Standard 523.6) Section 4.0, p. 29, first paragraph Design of Internal Dosimetry Programs” (HPS 2001a) Design of Internal Dosimetry Programs” (HPS 2011b) Section 4.3, p. 32, second paragraph a committed effective dose an E50 Section 4.3, p. 33, Table II Alternative Investigation Level] [0.1] [0.5] Section 4.4.6, p. 43, Performance Criteria for Radiobioassay” (HPS 1996a) Performance Criteria for Radiobioassay” (HPS 2011c) Section 4.5.6, p. 50 EPA 1992 EPA 2013 ASTM 1992 ASTM 2008 Section 5.1, p. 52, second paragraph If sufficient quantities of radionuclides are present or handled at a facility that accidental intakes resulting in 100-mrem E50 cannot be ruled out, an internal dosimetry program must be available

An internal dosimetry program must be available if sufficient quantities of radionuclides are present or handled at a facility in which accidental intakes resulting in 100-mrem E50 cannot be ruled out. Section 5.2, p. 52, first paragraph Baseline monitoring involves determining the worker’s bioassay status at the start of employment or potential exposure, and obtaining appropriate baseline measurements. Baseline monitoring involves determining appropriate baseline values for the worker at the start of employment or potential exposure. DOE-STD-1121-2008 v Table of Changes (cont’d) Section 6.0, p. 61, second paragraph that can be detected by bioassay Section 6.1, p. 61, second paragraph Due to the provisional acceptance of dose assessments based on workplace monitoring data, detailed methods are not described here. Where use of such data appears to be appropriate for dose assessment, the facility shall establish a protocol for their use as part of the internal dosimetry program, and document it in the technical basis documentation. Detailed methods are not described due to the provisional acceptance of dose assessments based on workplace monitoring data. The facility shall establish a protocol as part of the internal dosimetry program and document it in the technical basis documentation if use of workplace monitoring data appears to be appropriate for dose assessment. Section 6.2, p. 62 In fact, however, two distinct decisions are confounded by the current method: the first is the decision whether radioactivity above background levels is present, and the second is a decision whether any radioactivity that is present is above that which would be expected However, two distinct decisions are confounded by the current method: the first is the decision whether radioactivity is above background levels, and the second is a decision whether radioactivity is above what would be expected Section 7.3.2, p. 68 Delete: 1995c Section 7.5.6, p. 89 are dramatically different respectively Section 8.1, p. 95 Add: (see Section 5.2) Section 9.4, p. 104 Add: (Article 212) Section 10.3, p. 108 Delete: Lessons Learned Section 10.4, p. 110 Albert Wiley, MD, PhD albert.wiley@orise.orau.gov Carol Iddins, M.D. Carol.Iddins@orau.org Section 11.3, p. 112 Add: Chg 2 Add: Quality Assurance Section 12 Nonsubstantive updates to reflect current references mailto:albert.wiley@orise.orau.gov DOE-STD-1121-2008 i FOREWORD 1. This Department of Energy (DOE) standard is approved for use by all DOE Components and their contractors. 2. Constructive comments (recommendations, additions, deletions) and any pertinent data that may improve this document should be sent to Office of Worker Safety and Health Policy (EHSS-11) U.S. Department of Energy Washington, DC 20585 3. DOE technical standards, such as this standard, do not establish requirements. However, all or part of the provisions in a DOE standard can become requirements under the following circumstances: (1) they are explicitly stated to be requirements in a DOE requirements document; or (2) the organization makes a commitment to meet a standard in a contract or in an implementation plan or program plan required by a DOE requirements document.

Throughout this standard, the word "shall" is used to denote actions which must be performed if the objectives of this standard are to be met. If the provisions in this standard are made requirements through one of the two ways discussed above, then the "shall" statements would become requirements. It is not appropriate to consider that "should" statements would automatically be converted to "shall" statements as this action would violate the consensus process used to approve this standard. On June 8, 2007, DOE published an amendment to 10 CFR 835, Occupational Radiation Protection. This amendment updated DOE’s requirements for assessing and recording internal dosimetry results. The update was made to make DOE’s system consistent with more recent national and international consensus standards. DOE-STD-1121-2008 ii This page intentionally left blank. DOE-STD-1121-2008 CONTENTS LIST OF FIGURES ..................................................................................................................................... iv LIST OF EXAMPLES ................................................................................................................................. iv 1 INTRODUCTION............................................................................................................................... 1 1.1 SCOPE OF DOCUMENT ......................................................................................................... 1 1.2 PURPOSE ................................................................................................................................. 1 1.3 USE OF DOCUMENT .............................................................................................................. 2 1.4 OVERVIEW .............................................................................................................................. 2 1.5 USE OF NON-GOVERNMENTAL STANDARDS ................................................................ 4 2 DEFINITIONS AND ABBREVIATIONS ......................................................................................... 5 2.1 DEFINITION CROSS-REFERENCE ....................................................................................... 5 2.2 RADON AND THORON ....................................................................................................... 11 2.3 SPECIFIC DEFINITIONS ...................................................................................................... 12 2.4 ABBREVIATIONS, ACRONYMS, CODES, INITIALISMS, AND SYMBOLS ................. 17 2.5 CONVENTIONS FOR ROUNDING AND SIGNIFICANT FIGURES ................................ 19 3 DOCUMENTS AND PLANS LISTED BY THE IMPLEMENTATION GUIDE ........................... 20 3.1 INTERNAL DOSIMETRY TECHNICAL BASIS DOCUMENTATION ............................. 20 3.1.1 Organization and Agreements .................................................................................... 20 3.1.2 Bioassay Program Design .......................................................................................... 20 3.1.3 Participation in Bioassay Program ............................................................................. 21 3.1.4 Detection and Confirmation of Intakes ...................................................................... 22 3.1.5 Internal Dose Evaluation ............................................................................................ 23 3.1.6 Internal Dose Management ........................................................................................ 24 3.1.7 Records and Reporting ............................................................................................... 24 3.1.8 Medical Response ...................................................................................................... 25 3.1.9 Monitoring the Workplace ......................................................................................... 25

3.2 INTERNAL DOSIMETRY PROCEDURES MANUAL ....................................................... 25 3.2.1 Bioassay Contingency Plans ...................................................................................... 26 3.2.2 Dose Management Practices Plan .............................................................................. 27 3.2.3 Action Plan for Medical Response ............................................................................. 27 3.2.4 Quality Assurance Plan .............................................................................................. 28 4 DESIGN OF INDIVIDUAL MONITORING PROGRAMS FOR INTERNAL DOSIMETRY ...... 29 4.1 BIOASSAY COMPARED TO AND CONTRASTED WITH WORKPLACE AIR MONITORING ....................................................................................................................... 29 4.2 REFERENCE LEVELS AND DERIVED REFERENCE LEVELS ...................................... 32 4.3 INVESTIGATION LEVEL AND DERIVED INVESTIGATION LEVEL ........................... 32 4.4 DERIVED INVESTIGATION LEVELS ................................................................................ 33 4.4.1 Factors Affecting the DIL for Bioassay...................................................................... 33 4.4.2 Calculating the Derived Investigation Level for a Given Sample Frequency ............ 34 4.4.3 Factors Affecting the DIL for Air Sampling .............................................................. 39 4.4.4 Supplementing Routine Bioassay Programs When DIL < MDA ................................ 40 4.4.5 A Potential Technology Shortfall for Breathing Zone Air Sampling ......................... 40 4.4.6 Performance Specifications for a Bioassay Laboratory ............................................. 43 4.5 MEASUREMENTS OF WORKPLACE RADON AND THORON CONCENTRATIONS, POTENTIAL ALPHA ENERGY CONCENTRATIONS, AND MEASUREMENTS OF (OR ASSUMPTIONS ABOUT) EQUILIBRIUM FACTORS ....................................................... 48 4.5.1 Measurements ............................................................................................................. 48 4.5.2 Equilibrium Factors .................................................................................................... 48 DOE-STD-1121-2008 i 4.5.3 Performance Criteria for Instruments Used at DOE Sites to Characterize Airborne Radon and Thoron and Their Progeny ....................................................................... 49 4.5.4 Participation by DOE Sites in an Intercomparison Program for Radon Instrument Calibration, Precision, and Accuracy ......................................................................... 50 4.5.5 Calibration and Quality Control for Radon, Thoron, and Progeny Instrumentation .. 50 4.5.6 Use of Engineering Controls for Management of Exposures to Radon, Thoron, and Their Short-Lived Decay Products ............................................................................. 50 5 INDIVIDUAL MONITORING FOR INTERNAL DOSIMETRY .................................................. 52 5.1 SCOPE OF PARTICIPATION IN INDIVIDUAL MONITORING PROGRAMS FOR INTERNAL DOSIMETRY ..................................................................................................... 52 5.2 BASELINE INDIVIDUAL MONITORING: BIOASSAY .................................................... 52 5.3 PARTICIPATION IN ROUTINE INDIVIDUAL MONITORING PROGRAMS: BIOASSAY

AND/OR PERSONAL AIR SAMPLING ............................................................................... 53 5.3.1 Exposure Monitoring Thresholds for Radon and Thoron Progeny ............................ 56 5.4 SPECIAL BIOASSAY PROGRAM ....................................................................................... 56 5.5 TERMINATION AND ENDING-TASK BIOASSAY PARTICIPATION............................ 58 5.6 BIOASSAY FOR DECLARED PREGNANT WORKER ..................................................... 59 5.7 CONFIRMATORY BIOASSAY PROGRAM ....................................................................... 59 5.8 TIMELY RECEIPT OF BIOASSAY RESULTS ................................................................... 59 6 DETECTION AND CONFIRMATION OF INTAKES ................................................................... 61 6.1 USE OF WORKPLACE MONITORING DATA FOR DETECTING AND CONFIRMING INTAKES ................................................................................................................................ 61 6.2 USE OF BIOASSAY DATA FOR DETECTING AND CONFIRMING INTAKES ............ 62 6.2.1 Decisions Based on Individual Monitoring Data ....................................................... 63 6.3 STATISTICAL METHODS FOR CONFIRMING THAT AN INTAKE HAS OCCURRED 63 7 INTERNAL DOSE EVALUATION ................................................................................................ 64 7.1 DOSES TO BE ASSESSED ................................................................................................... 65 7.1.1 Committed Effective Dose ......................................................................................... 65 7.1.2 Committed Equivalent Dose to Tissue of Concern .................................................... 66 7.1.3 Total Effective Dose ................................................................................................... 66 7.1.4 Cumulative Total Effective Dose ............................................................................... 66 7.2 DATA NEEDS AND DEFAULT ASSUMPTIONS .............................................................. 66 7.3 INTERPRETATION OF BIOASSAY DATA ........................................................................ 67 7.3.1 Direct Estimation of Retained Quantity ..................................................................... 67 7.3.2 Biokinetic Modeling ................................................................................................... 67 7.3.3 Details of the Actual Dose Assessment ...................................................................... 70 7.3.4 Curve Fitting (Weighting of Data) ............................................................................. 70 7.4 CALCULATION OF INTERNAL DOSE FROM BIOASSAY DATA ................................. 74 7.4.1 Time or Time Course of Intake .................................................................................. 74 7.4.2 Intake and Dose Assessment for Mixtures of Radionuclides ..................................... 79 7.4.3 Special Considerations ............................................................................................... 80 7.5 CALCULATION OF INTERNAL DOSE FROM WORKPLACE DATA ............................ 80 7.5.1 Intake .......................................................................................................................... 81 7.5.2 Exposure in DAC-h .................................................................................................... 81 7.5.3 Assigned Respiratory Protection Factors for Use in Dose Evaluations ..................... 82 7.5.4 Assessment of Intake, Exposure, and Dose from Radon, Thoron, and Their Progeny

.................................................................................................................................... 82 DOE-STD-1121-2008 ii 7.5.5 Calculating Dose to Lung, and Intakes and Identities of Radon, Thoron and Their Progeny....................................................................................................................... 86 7.5.6 Possible Values of DACs for Pure Radon and Thoron Gas ........................................ 88 7.5.7 Choice of and Use of Assigned Protection Factors for Respirators in Radon and Thoron Dose Calculations .......................................................................................... 89 7.5.8 Determination of Radon and Thoron Background ..................................................... 90 7.5.9 Correcting for Relatively High Background PAECs .................................................. 91 7.6 SIMPLIFIED METHOD FOR DOSE ASSESSMENT FOR SMALL INTAKES ................. 91 7.7 UNCERTAINTIES ................................................................................................................. 91 7.7.1 Uncertainties Associated with Preliminary Evaluations ............................................ 94 7.7.2 Uncertainties Associated with Final Evaluations ....................................................... 94 8 INTERNAL DOSE MANAGEMENT .............................................................................................. 95 8.1 ROUTINE RADIOLOGICAL WORKER DOSE MANAGEMENT ..................................... 95 8.1.1 Management of Dose from Previous Intakes (Work Restrictions) ............................. 95 8.1.2 Compliance with Internal Dose Monitoring Requirements ........................................ 95 8.1.3 Control of Dose to the Embryo/fetus, Minors, and Students ..................................... 96 8.2 DOSE LIMITATION .............................................................................................................. 97 8.2.1 Interface and Coordination with the External Dosimetry Program and the Radiological Control Organization............................................................................. 97 8.2.2 Lifetime Dose Control ................................................................................................ 97 8.2.3 Doses Due to Intakes Prior to January 1, 1989 .......................................................... 97 8.2.4 Uncertainties ............................................................................................................... 97 8.3 DOSE CONTROL FOLLOWING ACCIDENTAL INTAKES ............................................. 98 8.3.1 Incident Dose Management ........................................................................................ 98 8.3.2 Preparation for Incidents Involving Intake ................................................................. 98 8.3.3 Internal Dose Control After an Incident ..................................................................... 99 9 RECORDS AND REPORTS .......................................................................................................... 101 9.1 WHAT TO RECORD - A GENERAL PHILOSOPHY OF RECORDS .............................. 101 9.2 REPORTING PRELIMINARY ASSESSMENTS OF UNPLANNED EXPOSURES ........ 102 9.3 PRECISION OF INTERNAL DOSE ASSESSMENTS ....................................................... 102 9.4 GUIDANCE ON LONG-TERM REEVALUATION OF INTAKES ................................... 103 9.5 GUIDANCE FOR PRACTICAL REPORTING OF INTERNAL DOSES .......................... 104 9.6 GUIDANCE ON CUMULATIVE TED ............................................................................... 104 9.7 RECORDS ASSOCIATED WITH BIOASSAY MEASUREMENTS AND THEIR

INTERPRETATION ............................................................................................................. 104 9.8 DOCUMENTING, RECORDING, AND RETAINING OF PAEC, PAEE, INTAKE, AND E50 FROM RADON AND THORON ......................................................................................... 105 10 MEDICAL RESPONSE .................................................................................................................. 106 10.1 NEED FOR MEDICAL RESPONSE ................................................................................... 106 10.2 ROLE OF THE HEALTH PHYSICIST IN MEDICAL TREATMENT .............................. 107 10.3 TREATMENT CRITERIA - WHEN TO TREAT ................................................................ 107 10.4 TREATMENT PROTOCOLS - HOW TO TREAT ............................................................. 108 10.5 IMPACT OF THERAPY ON DOSIMETRY ....................................................................... 110 10.6 COUNSELING WORKERS ................................................................................................. 111 11 QUALITY ASSURANCE .............................................................................................................. 112 11.1 GENERAL NEEDS .............................................................................................................. 112 11.2 INDEPENDENT REVIEW ................................................................................................... 112 11.3 COMPUTER SOFTWARE QUALITY ASSURANCE ....................................................... 112 11.3.1 Configuration Management ...................................................................................... 112 DOE-STD-1121-2008 iii 11.3.2 Verification and Validation (Acceptance) Testing of Codes.................................... 112 11.3.3 Corrections of Software Errors ................................................................................. 113 11.3.4 Software Security ..................................................................................................... 113 12 REFERENCES ................................................................................................................................ 114 APPENDIX A. REVIEW OF MEASUREMENTS OF EQUILIBRIUM FACTORS FOR RADON AND THORON PROGENY .................................................................................................................. 130 APPENDIX B. DEFINITIONS OF OBSOLETE OR REVISED QUANTITIES USED IN HISTORICAL RECORDS .................................................................................................................... 134 LIST OF TABLES Table I. Cross-Reference of Internal Dosimetry Terms ......................................................................... 5 Table II. Example Reference Level Magnitudes ................................................................................... 33 Table III. Acceptable Default Equilibrium Factors for Radon (FRn) .................................................. 49 Table IV. Summary of Numerical Conversions for Radon and Thoron Quantities, Regardless of the Precision of Measurements ......................................................................................... 88 Table V. Effective Dose Coefficients for Radon and Thoron Gas (Pure), Both Indoors and

Outdoors ............................................................................................................................. 89 Table VI. Three Options for Assigned Protection Factors for Rn, Tn, and Their Progeny ............. 90 Table VII. Default Background PAEC Values ....................................... Error! Bookmark not defined. Table VIII. Relative Importance of Various Sources of Uncertainty for Dose Assessment .............. 92 Table IX. Comparisons of Committed Effective Dose Detection Limits for Tritium Bioassay When 0.01 µCi/L of 3H Is Observed, as a Function of Time since Intake ............................... 93 Table X. Comparison of Methods of Assessing Dose from Intakes of Radionuclides ....................... 94 Table XI. Inhalation of Aged 6% Plutonium Mixture, No DTPA Given at Worksite ..................... 103 Table XII. Early Bioassay Measurement Results Corresponding to the Therapeutic Intervention Action Levels Used at the Hanford Site (Carbaugh 2007) (Part 1) ............................. 108 Table XIII. Early Bioassay Measurement Results Corresponding to the Therapeutic Intervention Action Levels Used at the Hanford Site (Carbaugh 2007) (Part 2) ............................. 109 Table XIV. Radon Equilibrium Factors Measured Outdoors ........................................................... 130 Table XV. Radon Equilibrium Factors Indoors at Home .................................................................. 131 Table XVI. Radon Equilibrium Factors Indoors at Work ................................................................. 132 Table XVII. Radon Equilibrium Factors in Uranium Mines ............................................................ 132 Table XVIII. Radon Equilibrium Factors in Non-Uranium Mines .................................................. 132 Table XIX. Thoron (220Rn) Equilibrium Factors ................................................................................ 133 DOE-STD-1121-2008 iv LIST OF FIGURES Figure 1. Alternative Logic Flow Chart for Determining the “Acceptable Minimum Detectable Dose” (AMDD) and DIL for Each Radionuclide or Group of Radionuclides When Exposure to Multiple, Independent Sources Is Possible ................................................ 36 Figure 2. Plot of the Normalized Detection Sensitivity as a Function of Number of Samples per Year for 3H ......................................................................................................................... 39 Figure 3. Reference Levels for Interpreting or Responding to Intake Monitoring Results .............. 62 Figure 4. Time Line for Intake Between Two Bioassay Measurements .............................................. 76 Figure 5. Expectation Value of Intake Divided by Intake at T/2 for a Single Exponential Retention Function .............................................................................................................................. 78 Figure 6. Expectation Value of Intake Divided by Intake at T/2 for a Single Exponential Retention Function .............................................................................................................................. 79 LIST OF EXAMPLES

Example 4.1. DIL for Type D Natural Uranium ................................................................................... 35 Example 4.2. Maximizing the Detection Sensitivity for Chronic Intakes of Tritium ........................ 38 Example 4.3. Use of Breathing Zone Air Samples to Supplement Routine Bioassay for Plutonium 42 Example 4.4. Improving Detection Capabilities of Air Sampling Using Averaging .......................... 42 Example 4.5. Potential Technology Shortfall for Breathing Zone Air Sampling of High Specific Activity Alpha Emitting Nuclides .................................................................................... 44 Example 4.6. The Number of Particles for Breathing Zone Air Sampling of a Lower Specific Activity Radionuclide ........................................................................................................ 45 Example 4.7. Example of Performance Specifications for a Bioassay Laboratory ............................ 46 Example 5.1. Baseline Bioassay Scenarios ............................................................................................. 53 Example 5.2. Criteria for Participating in Individual Monitoring Programs .................................... 54 Example 5.3. Circumstances Not Requiring Individual Monitoring................................................... 55 Example 5.4. Criteria for Commencing Special Bioassay .................................................................... 58 Example 7.1. Radionuclide Mixture: Sludge from Tanks Containing High Level Waste ................ 80 Example 7.2. Minimum Detectable Dose for a Nuclear Track Etch Radon Detector ....................... 86 Example 8.1. Dose Management Practices Regarding Internal Dosimetry ........................................ 96 Example 10.1. Situations Where Internal Dosimetry Actions and Medical Treatment .................. 106 DOE-STD-1121-2008 v This page intentionally left blank. DOE-STD-1121-2008 1 1 INTRODUCTION Internal dosimetry is “...the scientific methodology used to measure, calculate, estimate, assay, predict, and otherwise quantify the radiative energy absorbed by the ionization and excitation of atoms in human tissues as a result of the emission of energetic radiation by internally deposited radionuclides” (Raabe 1994). Radiation protection requirements for U.S. Department of Energy (DOE) and DOE - contractor employees are given in DOE’s Occupational Radiation Protection, Title 10, Code of Federal Regulations, Part 835 (DOE 2017). On June 8, 2007, DOE published an amendment to 10 CFR 835 which, in part, updated DOE’s requirements for assessing and recording internal dosimetry results. The update was made to make DOE’s system consistent with more recent national and international consensus standards. In this Technical Standard this regulation will be referred to as “10 CFR 835.” Further, the Radiological Control Standard (“RadCon Standard;” DOE-STD-1098-2017) contains provisions that apply to many contractors by virtue of being included in their contract. DOE’s 10 CFR 835 and RadCon Standard require monitoring of the workplace, and monitoring of radiation workers who, under typical conditions, are likely to receive 0.1 roentgen equivalent man (rem) (0.001 sievert (Sv)) or more committed effective dose, and/or 5 rem (0.05 Sv) committed equivalent dose to any organ or tissue, from all occupational radionuclide intakes in a year. The regulation 10 CFR 835 also requires that measurements of internal radionuclides and the assessments of committed effective dose resulting from intakes of radionuclides be recorded, reported, and archived.

1.1 SCOPE OF DOCUMENT This document applies to the internal dosimetry aspects of all Radiation Protection Programs of DOE and its contractors as required by 10 CFR 835.101 for the conduct of radiological work. As such, it provides detailed technical guidance on internal dosimetry to DOE and DOE-contractor personnel in fulfilling the requirements of 10 CFR 835 and applicable provisions of the RadCon Standard, as elaborated in the chapter, Internal Dosimetry Program, in the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE G 441.1-1C Chg 1) (DOE 2011a) by clarifying the requirements and providing specific examples of practical methods for conducting an effective internal dosimetry program. Guidance is provided on organization, staffing, training, and facilities; documents and plans; design of and participation in the bioassay program; internal dose evaluation; internal dose management; recording internal doses and related information; reporting of internal doses; medical response; quality assurance; and guidance for monitoring in the workplace as it applies to internal dosimetry. Details are provided on internal dosimetry aspects associated with radon, thoron, and their long-lived progeny; applications of models to bioassay data; dose assessment techniques; use of significant figures; and a guide to the wealth of internal dosimetry information at the various DOE sites. 1.2 PURPOSE This technical standard is created to provide a resource for those engaged in the science and practice of internal dosimetry within the DOE complex. This standard defines minimum levels of acceptable performance and provides basic procedural guidelines for evaluating the internal radiation equivalent and effective dose that may be received by radiation workers from intakes of radionuclides. This set of defined internal dosimetry performance criteria meets the requirements set forth in 10 CFR 835 for monitoring the workplace, for assessing internal radiation doses to workers at DOE facilities, and for recording and reporting requirements as they apply to internal dosimetry programs. DOE-STD-1121-2008 2 1.3 USE OF DOCUMENT This standard is for use in implementing the specific parts of the radiation protection programs required by 10 CFR 835.101 that relate to internal dosimetry programs. DOE and DOE-contractor personnel may use the specific methods and references in this standard as examples of acceptable means and methods to meet the internal dosimetry requirements of 10 CFR 835 and recommendations of the RadCon Standard, as elaborated in the chapter on Internal Dosimetry Program in the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a). 1.4 OVERVIEW Internal dosimetry is a major component of radiation protection for the approximately 80,000 radiation workers at DOE radiological or nuclear facilities. Workers who handle nuclear materials or who are involved in nuclear waste management are potentially at risk of inadvertent intakes of radioactive material. DOE policy and associated radiological control programs for limiting internal effective doses are based on containment of radioactive material to ensure (to the extent reasonably achievable) that radionuclides from work at radiological or nuclear facilities are not taken into the body. Most significant occupational intakes of radionuclides, i.e., individuals receiving doses approaching or exceeding a limit, occur as the result of contamination incidents associated with either the inadvertent release of radioactive material in the workplace or the unplanned loss of containment.

DOE’s 10 CFR 835 requires monitoring of employees likely to receive intakes of radionuclides that would result in committed effective doses at or above 100 mrem in a year. Monitoring programs in the workplace are designed to demonstrate that the requirements to limit exposure to 5 rem committed effective dose (E50) in any year are being met. Radiation worker bioassay monitoring programs are designed to provide the data needed to assess organ and tissue equivalent doses from intakes of radioactive material. If exposures to radioactive materials are such that significant internal doses are received from intakes occurring during the year, they are most often assessed using biokinetic models. In 1986, efforts were begun to develop a technical based manual that would provide guidance on developing and operating internal dosimetry programs at DOE radiological or nuclear facilities that would meet all applicable regulatory requirements. Input from internal dosimetry experts from DOE and various DOE contractors have been collected for two decades. This document, which resulted from that effort, attempts to assemble, in one place, information that will assist in meeting the requirements for conducting an internal dosimetry program within the DOE complex. The intent of this guidance document is to provide a fairly complete, though not exhaustive, set of basic procedural guidelines for achieving minimum levels of acceptable performance in evaluating the internal radiation equivalent and effective dose that may be received by radiation workers from intakes of radionuclides. The guidance provided here represents the collective wisdom of a diverse group with experience in internal dosimetry at DOE facilities. There has been a conscious effort to include examples from this group on the application of these guidance principles in the standard operations of their administered internal dosimetry programs. Section 2 provides the definitions and abbreviations that are commonly used in the field of internal dosimetry. Descriptions of documents and plans needed for an internal dosimetry program are provided in Section 3. These include internal dosimetry technical basis documentation, an internal dosimetry DOE-STD-1121-2008 3 procedures manual, a bioassay contingency plan for facilities having no routine monitoring program, a dose management practices plan, an action plan for medical response, and a quality assurance plan. Section 4 provides guidance on the design of an individual monitoring program. It gives specific information on the investigation level (IL), the derived investigation level (DIL), methods of measure- ment, frequency of bioassay measurement, supplementing routine bioassay programs (where the DIL < the MDA), and performance specifications for a bioassay or service laboratory. The different monitoring regimens of an individual monitoring program are discussed in Section 5. These include a baseline bioassay used prior to starting radiological work, routine bioassay monitoring conducted when workers are likely to receive 100 mrem committed effective dose in the workplace, special bioassay monitoring conducted following incidents with potential for intake, and bioassay monitoring conducted prior to termination of employment or end of potential for intake.

Section 6 contains the methods used to detect and confirm intakes of radioactive materials. The section explains the use of either bioassay data or workplace monitoring data to confirm an intake. Historically, workplace airborne radioactivity monitoring systems were put in place to detect inadvertent loss of containment. They were not intended to provide data for evaluating intakes by workers from exposures to airborne radioactivity. Thus, air monitors were located in areas with the highest potential for detecting loss of containment rather than in those areas most commonly occupied by radiation workers. Air monitoring data have not routinely been used to assess internal equivalent and effective dose because of the poor correlation between concentration of radionuclides in the air sampled by monitoring equipment and the actual amount of radioactive material inhaled by workers. While bioassay monitoring data are the preferred tool for intake and dose assessment, airborne exposure monitoring via DAC-hours tracking can have applications for chronic or acute low-level intakes, particularly when bioassay data cannot be obtained in a timely manner to allow adequate sensitivity to confirm or rule out a minor intake. Following the confirmation of an intake of radioactive material, an evaluation of the resultant internal dose is necessary. A discussion of the calculation of internal dose from bioassay data, and recommendations on interpretation of the bioassay data and handling of statistical uncertainties are presented in Section 7. Section 8 covers management of total effective dose (TED) and cumulative TED or lifetime occupational dose. Topics of discussion include routine occupational worker dose management, management of dose from previous intakes (work restrictions), compliance with internal dose monitoring requirements, control of dose to the embryo/fetus, minors, and students, and interface with external dosimetry. Guidance is provided on using and recording total effective dose, lifetime dose control, doses due to intakes prior to January 1, 1989, and statistical uncertainties. Also discussed are elements of an accidental dose control program, including incident dose management, preparation for incidents involving intakes, and internal dose control after an incident. Section 9 presents a discussion of recommendations for recording and reporting internal doses. Guidance is provided on a general philosophy of records and record keeping, reporting of preliminary assessments of unplanned exposures, precision of internal dose assessments, long-term reevaluation of intakes, practical reporting of internal doses, minimum recordable doses, recording of significant organ and tissue doses, cumulative TED, and records associated with bioassay measurements and their interpretation. Section 10 includes a recommended scheme for medical response following a potential intake of radioactive material. Guidance is provided on when and how to treat patients as well as the role of a DOE-STD-1121-2008 4 health physicist as an interface to medical treatment. The impact of therapeutic measures on the outcome of dosimetric evaluations is also discussed. Quality assurance issues associated with bioassay measurements, evaluations of intake, and internal dose are presented in Section 11. 1.5 USE OF NON-GOVERNMENTAL STANDARDS To the extent possible, this guidance document is written to be consistent with existing non- governmental standards for internal dosimetry, including:

ANSI N42.22-1995, American National Standard–Traceability of Radioactive Sources to the National Institute of Standards and Technology (NIST) and Associated Instrument Quality Control ANSI N42.23-1996, American National Standard–Measurement and Associated Instrumentation Quality Assurance for Radioassay Laboratories ANSI N323ab-2013, American National Standard for Radiation Protection Instrumentation Test and Calibration, Portable Survey Instruments ANSI Z88.2-2015, American National Standard for Respiratory Protection ANSI/HPS N13.1-2011, Guide to Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities ANSI/HPS N13.6-2010, Practice for Occupational Radiation Exposure Records Systems ANSI/HPS N13.12-2013, Surface and Volume Radioactivity Standards for Clearance ANSI/HPS N13.30-2011, Performance Criteria for Radiobioassay ANSI/HPS N13.39-2011, Design of Internal Dosimetry Programs DOE-STD-1121-2008 5 2 DEFINITIONS AND ABBREVIATIONS The definitions below come from many sources, indicated in the definition itself, and many have been adopted from the compilation used for the Internal Dosimetry Technical Basis Manual for the Mound Facility (Traub 1994). In this section, RadCon Standard refers to the U.S. Department of Energy Radiological Control Standard (DOE-STD-1098-2017); 10 CFR 835 refers to the DOE rule Occupational Radiation Protection (DOE 2017); AMG refers to the Air Monitoring chapter and IDG refers to the Internal Dosimetry Program chapter in the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a). Other definitions come from other DOE documents, national and international standards and recommendations; some definitions are new. Terms in italics are defined elsewhere in the definitions section. 2.1 DEFINITION CROSS-REFERENCE Most of the terms commonly used in the field of internal dosimetry have been adequately defined in documents that are commonly available at DOE sites and facilities. Rather than repeat the majority of these definitions here, Table I cross-references these definitions to other documents. Where a definition is found to have more than one source, the definition that occurs in 10 CFR 835 (when applicable) should be taken as the official definition for that term. Definitions are given in Section 2.3, when they are not given in 10 CFR 835, the RadCon Standard, IDG, or AMG, or when it is useful to present additional clarifying information. In Table I below, italicized items are used as symbols for the quantity elsewhere in this standard. Table I. Cross-Reference of Internal Dosimetry Terms Term = 10 CFR 835 RadCon Standard IDG AMG Other* activity median aerodynamic diameter (AMAD) X X activity median thermodynamic diameter (AMTD) ICRP-66 (1994a) administrative control level X X airborne radioactive material (or airborne radioactivity) X X airborne radioactivity area X X air monitoring X air sampling X ALARA Committee X alpha (α) (as a probability) X analyte X annual limit on exposure (ALE) ICRP-32 (1981) annual limit on intake (ALI) X X ICRP-32 (1981) for DOE-STD-1121-2008 6 Term = 10 CFR 835 RadCon Standard IDG AMG Other* 222Rn and 220Rn progeny appropriate blank ANSI/HPS N13.30-2011 assessment X assigned protection factor (APF) ANSI Z88.2-2015 as low as reasonably achievable (ALARA) X X background X synonymous with background radiation; ANSI/HPS N13.30-2011

background radiation X X baseline bioassay X becquerel (Bq) X beta (β) (as a probability) X Bias ANSI/HPS N13.30-2011 bioassay X X synonymous with radiobioassay breathing zone air monitoring X calibration X X Class SR-0 Gases, Class SR-1 Gases, Class SR-2 Gases ICRP Pub. 68 (1994a) company-issued clothing X confirmed intake X containment device X contamination area X X continuous air monitor (CAM) X X see “real time air monitoring” contractor X Via 10 CFR 820 controlled area X X counseling X critical mass X critique X decision level (DL, Lc) X declared pregnant worker X X DOE-STD-1121-2008 7 Term = 10 CFR 835 RadCon Standard IDG AMG Other* decontamination X derived air concentration (DAC) X X derived air concentration-hour (DAC-h) X X derived investigation level (DIL) X deterministic effects X Replaces nonstochastic effects diagnostic measurement ANSI/HPS N13.30-2011 direct (in vivo) radiobioassay X ANSI/HPS N13.30-2011 disintegration per minute (dpm) X DOE activity X X DOELAP X X dose X X absorbed dose (D) X X collective dose X committed effective dose (E50) X X committed equivalent dose (HT,50) X X cumulative total effective dose X X dose coefficient (hT(τ), e(τ), hT(50), e(50)) ICRP Pub. 68 (1994b) effective dose (E) X X equivalent dose (HT) X X equivalent dose to the extremity and skin X Replaces shallow dose equivalent dose to the lens of the eye X Replaces lens of the eye dose equivalent dose to the whole body X Replaces deep dose equivalent external dose or exposure X X internal dose or exposure X X radiation weighting factor (wR) X X Replaces quality factor tissue weighting factor (wT) X X total effective dose (TED) X X DOE-STD-1121-2008 8 Term = 10 CFR 835 RadCon Standard IDG AMG Other* whole body X X dose assessment X elimination X embryo/fetus X engineering controls X equilibrium factor (F) ICRP-65 (1993a) equilibrium equivalent concentration (EEC) ICRP-65 (1993a) evaluation X excretion X exposure X ICRP-65 (1993a) facility X false negative X false positive X fixed contamination X frisk or frisking X gastrointestinal (GI) tract model X ICRP-30 (1979a) ICRP-100 (2007) general employee X X gestation period X gray (Gy) X high contamination area X X high radiation area X X hot particle X hot spot X indirect (in vitro) bioassay X indirect radiobioassay ANSI/HPS N13.30-2011 individual X infrequent or first-time activities X intake X investigation level (IL) X in vitro measurement ANSI/HPS N13.30-2011 DOE-STD-1121-2008 9 Term = 10 CFR 835 RadCon Standard IDG AMG Other* in vivo measurement ANSI/HPS N13.30-2011 lifetime dose X lifetime occupational dose synonymous with lifetime dose lower limit on detection synonymous with MDA; ANSI/HPS N13.30-2011 member of the public X X minimum detectable amount (MDA) X ANSI/HPS N13.30-2011 minimum detectable concentration (MDC) ANSI/HPS N13.30-2011 minimum testing level (MTL) ANSI/HPS N13.30-2011 minor X X monitoring X X occupational dose X X person X personal air monitoring X personal protective equipment X personnel dosimeters X personnel monitoring X planned special exposure X potential alpha energy concentration (PAEC) ICRP-65 (1993a) potential alpha energy exposure (PAEE) ICRP-65 (1993a) prenatal radiation exposure X

protective clothing X qualification standard X quality assurance ANSI/HPS N13.30-2011 quality control ANSI/HPS N13.30-2011 rad X radiation X X radiation area X X radioactive material X radioactive material area X X DOE-STD-1121-2008 10 Term = 10 CFR 835 RadCon Standard IDG AMG Other* radioactivity X radiobioassay ANSI/HPS N13.30-2011 radiological area X X radiological buffer area (RBA) X radiological control hold point X radiological work X radiological work permit X radiological worker(s) X X radon X ICRP-65 (1993a) real time air monitoring X X X replacement for “continuous air monitoring” Reference Man X ICRP-23 (1975) ICRP-89 (2002) rem X removable contamination X respiratory protective equipment or device X X retention X routine radiobioassay monitoring X sievert (Sv) X site X screening measurements ANSI/HPS N13.30-2011 service laboratory ANSI/HPS N13.30-2011 sealed radioactive source X X special radiobioassay monitoring X state-of-the-art X step-off pad X sticky pad X stochastic effects X survey X technology shortfall X termination bioassay X DOE-STD-1121-2008 11 Term = 10 CFR 835 RadCon Standard IDG AMG Other* thermodynamic particle diameter (dth) ICRP-66 (1994a) thoron X ICRP-65 (1993a) Type I error X Type II error X Type F materials, Type M materials, Type S materials ICRP-66 (1994a) very high radiation area X X visitor X week X X whole body dose X working level (WL) X See 10 CFR 835 App A Footnote 4 year X X * Definitions whose source is other than 10 CFR 835, the RadCon Standard, IDG, AMG, or DOELAP are presented in Section 2.3. 2.2 RADON AND THORON The chemical element radon has two radiologically important isotopes that occur in nature: 220Rn and 222Rn. Following popular usage, this document refers to the former as “thoron” and the latter as “radon.” Radon and its short-lived progeny (decay products) are continuously produced by decay of 226Ra, a member of the naturally occurring 238U series. Airborne concentrations of radon’s short-lived progeny (218Po, 214Pb, 214Bi, and 214Po) are of interest due to their potential for deposition in the lung, leading to subsequent irradiation of lung tissue by alpha emissions from 218Po and 214Po. Thoron and its short-lived progeny are continuously produced by the decay of 224Ra, a member of the naturally occurring 232Th series. Thoron and 216Po have short half-lives: 56 s and 0.145 s, respectively. 212Pb and 212Bi are of interest due to the possibility of being deposited in the lung and irradiating tissue with alpha emissions from 212Bi and 212Po. DOE-STD-1121-2008 12 2.3 SPECIFIC DEFINITIONS Terms from 10 CFR 835 are used consistent with their regulatory definition. activity median thermodynamic diameter (AMTD): “Fifty percent of the activity (thermodynamically classified) in the aerosol is associated with particles of thermodynamic diameter (dth) greater than the AMTD. A lognormal distribution of particle sizes is usually assumed.” (ICRP 1994a) annual limit on exposure (ALE): The limit for potential alpha energy exposure to the progeny of 222Rn or 220Rn, expressed in units of working level months (WLM) (ICRP 1981b). An implicit ALE for other radionuclides is 2000 DAC-h. annual limit on intake (ALI): Reference 10 CFR 835.

Note: The ALI for 222Rn and 220Rn progeny is most correctly expressed in joules (J) of potential alpha energy (ICRP 1981b). Stochastic ALI (SALI) values and deterministic or nonstochastic ALI (NALI) values result from different dose limits. Intake of 1 SALI results in 5 rems committed effective dose, while intake of 1 NALI results in 50 rems committed equivalent dose to the most highly exposed tissue or organ. appropriate blank: A sample, person, or phantom that is, ideally, identical in physicochemically and radiologically significant ways with the sample, person, or phantom to be analyzed. (ANSI/HPS N13.30- 2011) assess: For purposes of this Standard, to officially assign or record a dose number. assigned protection factor (APF): The expected workplace level of respiratory protection that would be provided by a properly functioning respirator or a class of respirators to properly fitted and trained users. (ANSI Z88.2-2015) bias: A fixed mean deviation from the expected value that remains constant over replicated measurements within the statistical precision of the measurement. (Synonyms: deterministic error, fixed error, systematic error.) (ANSI/HPS N13.30-2011) bioassay: Another word for radiobioassay. biokinetic model: A series of often empirically determined mathematical relationships formulated to describe the intake, deposition in respiratory tract (if applicable), uptakes by the transfer compartment from intake compartment(s), uptakes by tissues or organs from the transfer compartment, translocation, retention, and elimination of a radionuclide from the body. censored data: Data that have been recorded as “less than” values rather than the observed numerical values (whether positive, zero, or negative). Class SR-0 gases: Insoluble and nonreactive gases and vapors (ICRP Publication 68, (1994a). Class SR-1 gases: Soluble or reactive gases and vapors (ICRP Publication 68, (1994a). Class SR-2 gases: Highly soluble or reactive gases and vapors (ICRP Publication 68, (1994a). DOE-STD-1121-2008 13 committed effective dose (E50): Reference 10 CFR 835. Note: For exposures to the short-lived radioactive progeny of 222Rn, committed effective dose is calculated directly from workplace measurements of potential alpha energy exposure using a dose conversion factor of 0.5 rems (0.005 Sv) per working level month (WLM). For exposures to the short-lived radioactive progeny of 220Rn, committed effective dose is calculated directly from workplace measurements of potential alpha energy exposure using a dose conversion factor of 1/6 rems (1/600 Sv) per WLM. Since the lung is the only tissue significantly irradiated by radon and thoron, the committed equivalent dose to lung due to exposures to radon and thoron is calculated by dividing the committed effective dose from radon and thoron by the tissue weighting factor for lung (wT = 0.12). committed equivalent dose ( HT,50): Reference 10 CFR 835. Note: For exposures to the short-lived radioactive progeny of 222Rn and 220Rn, see the definition of committed effective dose (below). compartment: The smallest element in a biokinetic model for which a mathematical representation of a retained quantity is given. Compartments may be organs (e.g., lung, liver), tissues (e.g., bone marrow), or systemic (e.g., the transfer compartment). critical level: Same as decision level. derived air concentration (DAC): Reference 10 CFR 835.

Note: The footnotes to Appendix A of 10 CFR 835 give some important information about the DACs listed. In particular, the right-hand column identifies the origin of each DAC–whether it was derived from the stochastic dose limit or the non-stochastic dose limit for a particular organ. Only DACs derived from the stochastic dose limit can be used to calculate committed effective dose directly from air sampling data. decision level (DL, LC): The number of counts measured or final instrument measurement of a quantity of analyte at or above which a decision is made that the analyte is definitely present. (ANSI/HPS N13.30- 2011) deposition fraction: The fraction of the amount of a material inhaled that is deposited in a particular region of the respiratory tract. For an aerosol, this fraction is a function of the aerodynamic or thermodynamic diameter. detection level (LD): This concept has been replaced by minimum detectable amount (MDA). diagnostic measurements: Measurements performed to estimate the amount of radionuclide deposited in a person when an intake is known or is suspected to have occurred. (ANSI/HPS N13.30-2011) direct radiobioassay: The measurements of radioactive material in the human body using instrumentation that detects radiation emitted from the radioactive material in the body (synonymous with in vivo measurement.). (ANSI/HPS N13.30-2011) DOE-STD-1121-2008 14 dose coefficient (hT(τ), e(τ),hT(50), e(50)): The committed equivalent dose per unit intake hT(τ) or committed effective dose per unit intake e(τ), where τ is the time period in years over which the dose is calculated (e.g., hT(50), e(50)) (adapted from ICRP Publication 68, p. vii). equilibrium factor (F): The equilibrium factor F with respect to potential alpha energy is the ratio of the equilibrium equivalent concentration (EEC) to the actual activity concentration of radon in air. equilibrium equivalent concentration (EEC): The EEC of a non-equilibrium mixture of short-lived radon progeny is that activity concentration of radon in radioactive equilibrium with its short-lived progeny that has the same potential alpha energy concentration as the non-equilibrium mixture to which the EEC refers. exposure: (1) The general condition of being subjected to radiation, such as by exposure to radiation from external sources or to radiation sources inside the body. In this document, exposure does not refer to the radiological physics concept of charge liberated per unit mass of air. (IDG) (2) The product of exposure time to a radioactive aerosol and the average concentration during exposure, divided by the value of the DAC for the radioactive material in question (expressed in DAC-h). (3) Exposure (of an individual to radon progeny) is the time integral of the potential alpha energy concentration in air over a given period (expressed in WLM) (adapted from ICRP Publication 65, p.4). gastrointestinal (GI) tract model: A mathematical representation of the behavior of radionuclides in the contents of the human gastrointestinal tract. indirect radiobioassay: Measurements to determine the presence of or to estimate the amount of radioactive material in the excreta or in other biological materials removed from the body (synonymous with in vitro measurement.) (ANSI/HPS N13.30-2011) intake compartment: One of four compartments from which systemic uptake can occur: the respiratory tract; the GI tract; a wound; or intact skin.

intake route: A pathway by which radioactive material enters the body. The main intake routes are inhalation, ingestion, absorption through the skin, and entry through injection or a cut or wound in the skin. in vitro measurement: Synonymous with indirect bioassay. in vivo measurement: Synonymous with direct bioassay. lower limit of detection (LLD): Synonymous with minimum detectable amount (MDA). minimum detectable amount (MDA): The smallest amount (activity or mass) of an analyte in a sample that will be detected with a probability ß of non-detection (Type II error) while accepting a probability α of erroneously deciding that a positive (non-zero) quantity of analyte is present in an appropriate blank sample (Type I error). (ANSI/HPS N13.30-2011) minimum detectable concentration (MDC): The minimum detectable amount (MDA) expressed in units of concentration. (ANSI/HPS N13.30-2011) DOE-STD-1121-2008 15 minimum detectable (effective) dose: The minimum detectable (effective) dose associated with a bioassay program. Formerly called “missed dose.” minimum testing level (MTL): The minimum amount of radioactive material that the testing laboratory should use to estimate relative bias and relative precision of the service laboratories participating in the performance testing program, assuming the sample(s) is(are) free of interference from other radionuclides unless specifically addressed. The MTL is a multiple of the expected procedural MDA. (ANSI/HPS N13.30-2011) potential alpha energy concentration (PAEC): The kinetic energy potentially released in a unit volume of air by alpha particles emitted by the short-lived radioactive progeny of 222Rn (i.e., 218Po and 214Po) or 220Rn (i.e., 216Po, 212Bi, and 212Po). PAEC is expressed in working levels (WL). potential alpha energy exposure (PAEE): The average potential alpha energy concentration (PAEC) to which a worker is exposed, multiplied by the time of exposure in working months of 170 hours: that is, PAEE = PAEC × time. PAEE is expressed in working level months (WLM). quality assurance: All those planned and systematic actions necessary to provide adequate confidence that an analysis, measurement, or surveillance program will perform satisfactorily in service. (ANSI/HPS N13.30-2011) quality control: Those actions that control the attributes of the analytical process, standards, reagents, measurement equipment, components, system, or facility according to predetermined quality requirements. (HPS N13.30-2011) radiobioassay: Measurement of the amount or concentration of radionuclide material in the body or in biological material excreted or removed from the body and analyzed for purposes of estimating the quantity of radioactive material in the body. (ANSI/HPS N13.30-2011) radon: For purposes of this DOE Standard, unless otherwise specified, the isotope 222Rn. respiratory tract model: A mathematical representation of the behavior of particles and gases in the human respiratory tract. retained quantity: The amount of material which, after being taken into the body by inhalation, ingestion, entry through an open wound, or absorption through the skin, exists in the whole body, a compartment, an organ, or a tissue at a specified time. screening measurements: Measurements made to detect radioactive material under routine conditions but not used to quantify the amount of a given radionuclide. (ANSI/HPS N13.30-2011)

service laboratory: Laboratory performing direct and/or indirect radiobioassay measurements. (ANSI/HPS N13.30-2011) thermodynamic particle diameter (dth): Diameter (in µm) of a spherical particle that has the same diffusion coefficient in air as the particle of interest. (ICRP 1994a) thoron: The isotope 220Rn, also symbolized by Tn. Thoron is a “conventional name” like tritium. DOE-STD-1121-2008 16 translocation: Movement within the body of a radioactive material after uptake, such as from bone to kidney. working level (WL): is any combination of the short-lived radioactive progeny in one liter of air, without regard to the degree of equilibrium, that will result in the ultimate emission of 130,000 MeV of alpha energy. (Footnote to Appendix A, 10 CFR 835) Note: WL is the unit of potential alpha energy concentration (PAEC) (1 WL = 2.083 E-5 J/m3). working level month (WLM): The unit of potential alpha energy exposure (PAEE), defined as exposure for 1 working month (of 170 hours) to an airborne concentration of 1 WL. (1 WLM = 1 WL × 170 hours = 0.00354 Jּh/m3). wound compartment: The compartment in a biokinetic model whose retained quantity is the amount of radioactive material in a wound that has not moved to the transfer compartment. visitor: A member of the public entering a controlled area. DOE-STD-1121-2008 17 2.4 ABBREVIATIONS, ACRONYMS, CODES, INITIALISMS, AND SYMBOLS α Alpha β Beta ∆Amin detection sensitivity µ prefix micro (10-6) ALARA as low as reasonably achievable ALE annual limit on exposure ALI annual limit on intake AMAD activity median aerodynamic diameter AMDD Acceptable minimum detectable dose AMG Air Monitoring Chapter of 10 CFR 835 Implementation Guide AMTD activity median thermodynamic diameter ANSI American National Standards Institute APF assigned protection factor ASTM American Society for Testing and Materials BEIR Biological Effects of Ionizing Radiation BZ breathing zone CAM continuous air monitor CFR Code of Federal Regulations Cu observed concentration of analyte in urine D absorbed dose DAC derived air concentration DIL derived investigation level DL decision level DOE U.S. Department of Energy DOELAP DOE Laboratory Accreditation Program dpm disintegration per minute DRL derived reference level dth thermodynamic diameter DTPA diethylene-triamine-pentaacetic acid EDTA ethylene-diamine-tetraacetic acid EEC equilibrium equivalent concentration EEI equilibrium equivalent intake EPA U.S. Environmental Protection Agency f sampling frequency F equilibrium factor FEMP Fernald Environmental Management Project fp unattached fraction FR Federal Register GA general area GI gastrointestinal GM Geiger-Müller GSD geometric standard deviation DOE-STD-1121-2008 18 Gy Gray E effective dose e(50) 50-year effective dose coefficient (committed effective dose per unit of activity taken in) eing(50) 50-year effective dose coefficient for ingestion (committed effective dose per unit of activity ingested) einh(50) 50-year effective dose coefficient for inhalation (committed effective dose per unit of activity

inhaled) HT,50 committed equivalent dose HPS Health Physics Society IDG Internal Dosimetry Chapter of 10 CFR 835 Implementation Guide ICRP International Commission on Radiological Protection IL investigation level ILs sample investigation level IRF intake retention fraction IRFu intake retention function for urinary excretion ISO International Organization for Standardization LC decision level (formerly the critical level) LD detection level (use MDA) LMR medical referral level (LDMR is the derived medical referral level) LV verification level (LDV is the derived verification level) LLD lower limit of detection MDA minimum detectable amount (or activity) MDC minimum detectable concentration MLE maximum likelihood estimator MPBB maximum permissible body burden MTL minimum testing level MWA maximum working activity NALI nonstochastic or deterministic annual limit on intake NCRP National Council on Radiation Protection and Measurements NIOSH National Institute of Occupational Safety and Health NIST National Institute of Standards and Technology NRC Nuclear Regulatory Commission NV the number of transitions per unit volume PAEC potential alpha energy concentration PAEE potential alpha energy exposure QA quality assurance QC quality control RBA radiological buffer area RWP radiological work permit SALI stochastic annual limit on intake SI International System (of units) t time t0, t1, t2 particular values of time tE exposure time (d) TED total effective dose TLD thermoluminescent dosimeter Tn thoron (220Rn) DOE-STD-1121-2008 19 τ time since intake (lower case Greek letter tau) V&V verification and validation V u urine excretion rate UMTRA Uranium Mill Tailings Remedial Action UNSCEAR United Nations Scientific Committee on the Effects of Atomic Radiation WL working level WLM working level month wR radiation weighting factor wT tissue weighting factor 2.5 CONVENTIONS FOR ROUNDING AND SIGNIFICANT FIGURES The need to distinguish between exact numbers, such as dose limits, and imprecise numbers, such as the results of measurements, leads to the use of some conventions in this DOE Standard. Any legislated number, as well as any integer or ratio of two integers, is an exact number with no uncertainty1. Examples of exact numbers are the 5-rem annual TED limit, the DAC for radon progeny of 0.83 WL, and integral fractions and exponents (e.g., kinetic energy = ½ mv2). Exact numbers may have tolerances, but when tolerances are not specified, the exact numbers must be treated as arbitrarily precise: a 5-rem limit is 5.000 000 000 rem. Measurements are often uncertain and imprecise, and inferences of dose from measurements using calculational models with uncertain parameters are also uncertain and imprecise. Confusion sometimes results when comparing uncertain or imprecise numbers with exact standards. Furthermore, difficulty arises when exact numbers are converted from one set of units to another, and the result is rounded. This difficulty becomes particularly acute for quantities and units associated with radon and thoron. Thus, the DOE has decided to derive all radon and thoron concentration values from 10 CFR 835, Appendix A, PAEC limits (or ICRP/IAEA PAEE limits for newer recommendations), rather than from 10 CFR 835, Appendix A, equilibrium equivalent DACs, which give slightly different answers and lead to confusion (Strom et al. 1996).

Excellent, detailed guidance on significant figures and rounding for measurements is given by the ASTM (ASTM 1993). Unfortunately, ASTM E380-93 does not recognize the exact nature of regulatory limits nor address the problems of significant figures when converting exact numbers between unit systems. Also, it does not address radon and thoron quantities and units. To minimize roundoff errors, it is recommended that all calculations be performed using numbers specified to at least “single precision” (six to seven significant figures) or as rational numbers if appropriate (ratios of integers, e.g., 1/3, 5/12, etc). For reporting purposes, it should be acceptable to round to three significant figures or to the precision of the reporting field, whichever is less. For example, 3.84 mrem may be rounded to 4 mrem if only integral numbers of mrem can be reported in a given field. More detail on recording and reporting is given in Section 9. 1Some irrational numbers are exact, such as π, √2, and e, the base of the natural logarithms. Sometimes, unit conversions are exact, such as 37,000,000,000 Bq/Ci. DOE-STD-1121-2008 20 3 DOCUMENTS AND PLANS LISTED BY THE IMPLEMENTATION GUIDE This section provides suggested contents for the various documents listed in the IDG and the RadCon Standard to provide technical guidance for implementing internal dosimetry programs. Numerous organizational and hierarchal approaches to implementing documents are possible. It is not the intent of this technical standard to advocate any specific approach. However, the topics and items addressed in this section are considered extremely important to a technically and administratively functional internal dosimetry program. Internal dosimetry documents and plans shall be rooted in the requirements of 10 CFR 835 and other contractual requirements, for example the RadCon Standard. They should draw guidance from the IDG, this technical standard and applicable non-government standards and draft standards, such as those listed in Section 1.5. 3.1 INTERNAL DOSIMETRY TECHNICAL BASIS DOCUMENTATION This section summarizes all of the topics that appear throughout the IDG and the RadCon Standard (DOE-STD-1098-2017) for the Internal Dosimetry Technical Basis Documentation. The RadCon Standard recommends the development of Internal Dosimetry Technical Basis Documentation that gives scientific information and other rationale explaining each element of the internal dosimetry program to support dose evaluation methods used therein. The technical basis documentation should be reviewed periodically and updated as necessary to ensure that the scientific bases are appropriate for current conditions. The following information is suggested for inclusion in various sections of the internal dosimetry technical basis documentation: 3.1.1 Organization and Agreements • letter(s) of agreement between contractors at a multiple-contractor site detailing the responsibilities, authority, and communication needs of the respective parties • listing of arrangements between the internal dosimetry program and the bioassay measurements laboratory, including • needed turnaround times • MDAs for special and routine samples • priorities for classification of samples (e.g., routine, special, emergency) 3.1.2 Bioassay Program Design • physical and chemical characteristics of radioactive materials encountered in the workplace • establishment of the type and frequency of measurements to be used (RadCon Standard 522.4)

DOE-STD-1121-2008 21 • derivation of decision levels • default trigger levels • preliminary actions to be taken for exposures to the different radionuclides present at a facility following suspected or confirmed intakes at various levels • tailoring of investigations to a specific individual worker or exposure circumstances • documentation of the derivation of DILs • established DILs for each bioassay method applied for the analysis of all radionuclides to which workers are likely to be exposed • if it is known or is likely that an individual has or could have intakes during the year from different sources that could result in doses above the IL, methods to use to derive appropriately smaller DILs • methods of bioassay measurement and the rationale or justification for each • the MDAs for the bioassays • justification for the bioassay monitoring frequencies, including an evaluation of the largest internal dose (i.e., minimum detectable dose) from an intake (acute or chronic) that could go undetected with the chosen frequency • documentation and justification of a planned supplementary approach for intake or dose assessment in case of technology shortfall • the rationale for the formal action procedures following a bioassay result unexpectedly above the DL 3.1.3 Participation in Bioassay Program • rationale for selection of workers for bioassay monitoring For the purpose of compliance with 10 CFR 835.402(c)(l), all sources of occupational intakes must be included in making the determination that an individual is likely to receive a committed effective dose (committed effective dose) of 100 millirem or more in a year. For example, if a determination is made that an individual was likely to receive a committed effective dose of 95 millirem in a year from one radionuclide and this individual was also likely to receive a committed effective dose of 10 millirem in a year from another radionuclide, then that individual would need to be monitored in accordance with 10 CFR 835.402(c)( 1). After the determination is made that an individual needs to be included in the internal dosimetry program, the organization responsible for compliance with 10 CFR 835 must assess the anticipated magnitude of the occupational intakes from radionuclides to which the individual will be exposed and the feasibility and cost associated with monitoring at the anticipated exposure levels. At a minimum, the internal dosimetry program shall include evaluation of dose from all radionuclides contributing significantly to the individual’s dose. The basis for including or excluding certain exposures DOE-STD-1121-2008 22 to radionuclides in internal dosimetry programs should be documented in the site’s Internal Dosimetry Technical Basis Document. Generally, it is acceptable for an individual to be no longer monitored as part of the internal dosimetry program if that individual’s work conditions have changed such that they are not likely, under typical conditions, to receive a committed effective dose of 100 millirem or more during the remainder of the year. However, per 835.402(d), internal dose monitoring programs implemented to demonstrate compliance with 835.402(c) need to be adequate to demonstrate compliance with the dose limits. Individuals who have already been assessed a dose approaching the dose limits may need to have continued monitoring to meet this requirement. The decision for continued participation in the internal dosimetry program will require

consideration of the individual’s current dose, the anticipated dose for the remainder of the year, and types of radionuclides. If an individual will likely receive no additional dose for the rest of the year, or if the additional internal dose will likely still result in a total internal dose of less than 100 millirem in a year, continued internal dose monitoring is not required. Other situations should be assessed on a case by case basis. Chapter 5 gives examples of criteria for participation in a bioassay program, including guidance on ending-task bioassay participation and participation in routine bioassay programs if respiratory protection is used to limit intakes of radioactivity. The basis for continuing or terminating individual participation in internal dosimetry programs shall be documented in the site’s Internal Dosimetry Technical Basis Document. 3.1.4 Detection and Confirmation of Intakes • biokinetic models • model parameters • assumptions • justification of the choices of default parameters used in deriving a DIL • parameters and their associated default values used in dosimetric modeling and evaluation, such as • intake date • deposition probabilities (deposition fractions) • retention functions • organ masses • absorption fractions • facility-specific factor DOE-STD-1121-2008 23 • statistical methods for • evaluating bioassay data • identifying bioassay results above environmental background values • using appropriate blanks for analyzing trends • MDAs • description of a procedure for evaluating doses if the time course of an intake cannot be plausibly established • if DAC-h calculations are used to assess exposures to airborne radioactive materials, a description of any authorized adjustment(s) to such calculations to account for the use of respiratory protection 3.1.5 Internal Dose Evaluation • method for evaluating internal doses from routine and special bioassay data, and where appropriate, from workplace monitoring data, including personal air samplers • methods for calculating internal doses • methods for evaluating equivalent doses from specific radionuclides, mixtures of radionuclides, and materials of differing chemical characteristics • basis for the evaluation methods including recommendations given in ICRP Publications and NCRP Reports, which embody improvements and updates of the science of internal dosimetry • justification for alternative approaches and assumptions used in dose calculations • dose evaluation quality assurance • biokinetic models • model parameters • assumptions • the basis for determining which individual-specific and facility-specific factors are expected to change the dose calculations by a factor of 1.5 or more • a description of the level of intake or committed effective dose detection achieved • a basis for projecting a committed effective dose of one IL from bioassay results (i.e., how are multiple bioassay results over the course of a year evaluated and compared to the IL) DOE-STD-1121-2008 24 3.1.6 Internal Dose Management • action levels for administrative response to intakes of radionuclides by workers, including decisions reached among medical, management, and radiation protection staff • a description of the site policy for confirming intakes in instances of historical bioassay data prior to January 1, 1989, where follow-up bioassay samples were not required on positive bioassay samples or where documentation is lacking (counter efficiency, chemical recovery, minimum detectable amount/activity, etc.)

• methodology to account for the portion of a bioassay result that may be due to one or more prior confirmed intakes • basis for work restrictions used during internal dose evaluation • administrative controls to limit dose to declared pregnant workers, minors, and students • description of the interface with external dosimetry • methods for calculating TED • determination of lifetime dose and specification of lifetime dose administrative control levels 3.1.7 Records and Reporting • methods for documenting calculations • recording and reporting practices for internal dosimetry • a description of the configuration control of the internal dosimetry technical basis documentation, including • specific maintenance of the internal dosimetry technical basis documentation, including responsibilities for authorship, review, approval, and distribution • maintenance as a controlled document • periodic review of internal dosimetry technical basis documentation by the site radiation protection organization to ensure that the scientific bases are current, and that the technical basis appropriately reflects changes in existing standards, anticipated changes, and new standards • external peer-review by qualified individuals on a periodic basis • retention of radiological protection program records with copies of all previous revisions and changes retained for future program review DOE-STD-1121-2008 25 3.1.8 Medical Response • description of accidental dose control methods 3.1.9 Monitoring the Workplace • specification of continuous air monitor (CAM) alarm levels and justification of the levels chosen 3.2 INTERNAL DOSIMETRY PROCEDURES MANUAL Written policies and procedures covering each step in the activities used to determine worker internal dose are an essential element of an acceptable internal dosimetry program. All elements of the internal dosimetry program shall be specified in written procedures. These procedures should be consistent with 10 CFR 835, the RadCon Standard, the IDG, relevant DOE Orders, this document, and the internal dosimetry technical basis documentation. The internal dosimetry procedures shall specify or identify the following: • methods and requirements for measurement (bioassay) and evaluating and recording internal dose • methods for consistent collection of workplace and personnel monitoring data, its evaluation, documentation of results, and records maintenance • the components of the internal dosimetry program and the organizational structure to which it reports • responsibilities of line management and members of the dose evaluation group • elements of the workplace and radiological worker monitoring programs that are germane to internal dosimetry • guidelines for prompt follow-up of worker intakes of radioactive materials, and appropriate follow-up response to intakes, including the medical management of workers with excessive intakes • all relevant subcontractor procedures to be included in the historical record files of the DOE contractor • the MDAs of the various bioassay measurement methods • programmatic details, including: - method(s) of bioassay measurements (e.g., urinalysis, fecal analysis, or in vivo counting) - analytical methodology (e.g., chemical separation followed by alpha counting) - measurement parameters (e.g., counting time or instrument efficiency) to be used in each component of the bioassay program DOE-STD-1121-2008 26 • frequency of the routine bioassay program

• agreements with the bioassay measurements laboratory on needed turnaround times, MDAs for special and routine samples, and priorities for classification of samples (e.g., routine, special, emergency) • factors to be considered by the internal dosimetry staff in determining the follow-up or confirmatory actions to be taken in response to positive bioassay results • actions taken following a bioassay result unexpectedly above the DL • personnel who will establish confirmatory bioassay requirements in cases not covered by the procedures • trigger levels and preliminary actions to be taken for exposures to the different radionuclides encountered at the facility • other methods that may be used for the evaluation of doses from intakes and their scientific basis • action levels for administrative response to intakes of radionuclides by workers • records to document the appropriateness, quality, and accuracy of monitoring methods, techniques, and procedures in use during any given period, pursuant to applicable requirements and standards • documentation that all steps in the activities that control or evaluate worker internal doses by written procedures provide appropriate quality control and quality assurance Radiochemical laboratories and in-vivo counting facilities whose measurements are used by internal dosimetry programs shall have written procedures that can be referenced by internal dosimetry programs. The internal dosimetry program shall receive periodic assessment by the site radiation protection organization to review dose assessment procedures as necessary to ensure that the program maintains the capability to stay abreast of scientific developments in internal dosimetry and provides a quality radiation protection service to workers. Paragraph 10 CFR 835.102 requires that an internal audit be done every 36 months. External peer-review by qualified individuals on a periodic basis is also recommended. The procedures shall be reviewed at least once every two years and updated as necessary. The needs for maintenance of procedures shall be specified, including responsibilities for authorship, review, approval, and distribution. 3.2.1 Bioassay Contingency Plans Some facilities with low potential for significant occupational intake of radioactivity may not have any routine bioassay program. Examples of such facilities are those where only sealed sources are handled, or the types, quantities, and frequency of dealing with radioactive materials does not support establishment of routine capability from a cost-effectiveness viewpoint. However, if quantities of unsealed radioactive material are handled infrequently or if accidents could happen causing intakes DOE-STD-1121-2008 27 corresponding to 100 mrem committed effective dose, then it may be wise to have a contingency plan for obtaining bioassay measurements. Elaborate advance arrangements are not necessarily warranted. However, thought should be given to what types of bioassay measurements might be needed, and how and where they would be obtained. A good approach would be to identify the closest DOE facility with capability appropriate for the radionuclides and to have a letter of agreement or memorandum of understanding in place to obtain measurements on an as-needed basis. As a minimum, the radiation protection organization should know who to contact for support, how long until data could be obtained, and what to do until data would become available.

A contingency plan for sites having routine bioassay is worth considering because of the possibility of losing one or more components of a bioassay program. Such loss could result from equipment or facility failure, or from loss of vendor services. For instance, a site that relies on a contracted laboratory for radiochemistry analysis for bioassay samples could suddenly find itself in a crisis if the contracted laboratory were to close or the contract were canceled. Without timely support and re-establishing capabilities, site operations could be significantly impaired. A contingency plan with another DOE site to provide some limited, short-term support could allow normal site operations to continue. The intent of this discussion of bioassay contingency plans is not to recommend establishment of a formally documented plan with implementing procedures, but that some clear thought be given to appropriate actions. The actual “plan” may be simply a paragraph or subsection in the technical basis manual or procedures identifying the contact point at another site for such support, and some indication of what would be needed (for example, a contract or inter-contractor order) to begin the support. A documented letter of agreement or understanding would be desirable. 3.2.2 Dose Management Practices Plan Since DOE’s radiation protection program is based on total effective dose, dose management requires coordination between a site’s internal dosimetry program and its external dosimetry program. For example, during an evaluation of an internal dose case, it may be important to restrict a worker’s external dose. Similarly, if lifetime dose controls (as given in the RadCon Standard) were being used, a periodic reassessment of the internal doses could influence lifetime occupational dose decisions. The dose management practices plan may be a part of the internal dosimetry procedures manual. Alternatively, the plan may be part of a higher echelon manual or contained in external dosimetry or other procedures. 3.2.3 Action Plan for Medical Response This plan shall describe the coordinated response when a medical injury is combined with potential internal dose concerns or when an intake may be sufficiently large to warrant therapeutic medical intervention for dose reduction. Medical response requires coordination between the radiation protection and medical organizations. The coordination can become even more complex when multiple contractors or subcontractors are involved and in situations where some medical services are provided by onsite personnel, and some are provided by offsite sources. For example, onsite services usually include some kind of first aid response and may even involve nursing and medical doctor or physician’s assistant staff. At the same time, emergency medical services (ambulance and medical trauma support) may be provided by offsite private or public organizations. A clear understanding and delineation of responsibilities and authorities in the treatment of contaminated injuries or for dose reduction therapy shall be included in the action plan. This medical response action plan may be part of the internal dosimetry procedures or an element of other site documents. DOE-STD-1121-2008 28 Some examples of combined medical response and internal dose concern scenarios are provided in Section 10. Technical guidance for internal dosimetry efforts in support of medical response is also provided in Section 10.

3.2.4 Quality Assurance Plan All steps in the activities that control or evaluate worker internal doses shall be covered by written procedures that provide appropriate quality control and quality assurance. The quality assurance plan may be a section in technical basis documentation. More information is provided in Section 11. DOE-STD-1121-2008 29 4 DESIGN OF INDIVIDUAL MONITORING PROGRAMS FOR INTERNAL DOSIMETRY In the context of internal dosimetry, individual monitoring includes routine bioassay (mentioned in 835.402(c)) and/or personal air sampling (not mentioned in 835.402(c)). The chapter on Internal Dosimetry Program in the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a) provides general guidance for the design of a bioassay program, but little guidance for air monitoring programs as a basis for internal dosimetry. In addition to considering all points in the IDG, DOE sites should strive to comply with ANSI/HPS N13.39, “Design of Internal Dosimetry Programs” (HPS 2011b) when that standard is not in conflict with 10 CFR 835, the RadCon Standard, and the IDG, as appropriate. One conflict between ANSI/HPS N13.39 and the IDG is in the definition of investigation level (see Section 4.2). In this conflict the IDG shall prevail. There is a wealth of information on design of bioassay programs in the technical basis documentation of many DOE sites (Carbaugh et al. 2003a; Hill and Strom 1993; Traub 1994; Baker et al. 1994; Inkret and Miller 1995; Calvo and McLaughlin 1995; Crandall et al. 2001). The reader is advised to consult this documentation for details of bioassay program design. Additional useful information on design can be found in works by Skrable (Skrable 1992); in element-specific standards (HPS 1996e; HPS 1994), in Regulatory Guides of the NRC (NRC 1992a, 1993a), in works of the ICRP (ICRP 1988) and NCRP (NCRP 1985a), the Y-12 Internal Dosimetry Technical Basis Document (Snapp 2007) and in the Hanford Internal Dosimetry Program Manual (Carbaugh 2009). Less information is available on design of personal air sampling programs as a basis for internal dosimetry. Readers should consult the Air Monitoring chapter of the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a) and documents of the NRC (NRC 1992a, 1992b, 1993a; Hickey et al. 1993). Guidance is given below on individual monitoring for the short-lived progeny of radon and thoron. As used in this DOE Standard, personal air monitoring refers to assigning specific air monitoring results to individual workers, regardless of whether the air monitoring was accomplished by general area sampling, breathing zone sampling, or individual personal (lapel) air samplers. 4.1 BIOASSAY COMPARED TO AND CONTRASTED WITH WORKPLACE AIR MONITORING DOE’s occupational radiation protection system is dose-based. 10 CFR 835.209(b) is the only requirement that addresses methods of internal dose assessment: The estimation of internal dose shall be based on bioassay data rather than air concentration values unless bioassay data are: (1) unavailable; (2) inadequate; or (3) internal dose estimates based on air concentration values are demonstrated to be as or more accurate. 10 CFR 835.209(b) does not require sites to use air monitoring data for internal dose assessment, but permits sites to use air monitoring data under certain conditions. “Inadequate bioassay,” for compliance with 10 CFR 835.209(c), may be taken to pertain to radionuclides with effective half-lives too short to be feasible for routine or special bioassay. Such radionuclides include radon and thoron and their

DOE-STD-1121-2008 30 short-lived progeny, as well as radionuclides such as 227Th, 223Ra, 225Ra, and 225Ac when separated from their long-lived parents. It may also include routine bioassay for isotopes with high dose per intake coefficients, such as many of the transuranics. As defined in the IDG, a technology shortfall, such as for routine bioassay monitoring for Pu, does not preclude the use of routine bioassay monitoring nor force the use of air sample data for dose calculations. Rather, the IDG suggests that the capabilities of the bioassay program be stretched as far as reasonable, that workplace monitoring be enhanced, and that state-of-the-art techniques be used in general. Reliance should be placed on prompt detection of possible intakes in the workplace, and that special bioassay should be promptly initiated (usually the same day) when intakes are suspected. In vivo count times should be as long as reasonable, and MDAs should be as low as reasonably achievable, with an emphasis in both cases on “reasonable” as explained in the IDG. Air sample data may be used for initiating special bioassay without being used for dose assessment. When technology shortfalls occur, the enhancements recommended in the IDG provide an acceptable method for complying with 10 CFR 835. By performing air sampling and documenting the results, in combination with an effective access control program, worker exposure measured in DAC-h can be tracked. Internal dosimetry programs typically base bioassay frequency and type on levels of actual or anticipated exposures to individuals. By tracking DAC-h for individuals, the type and frequency of needed bioassay measurements can be determined. For example, if a radiological worker receives less than 40 DAC-h (2 percent of an ALI) in a year with no respiratory protection, the individual would not be scheduled to participate in the bioassay monitoring program for that year. Additionally, participation of the individual in the bioassay monitoring program for the next year should be considered. In the case where bioassay measurements may not be available, or their validity is questionable, internal dose assessments can be determined from the number of DAC-h tracked for that individual. When DAC-h are used for this purpose, any adjustments, such as protection factors for respiratory protection, must be documented. Air sampling and monitoring play an integral role in dose assessment for all isotopes, including those where the DIL is less than the detection capability. By tracking DAC-h, the expected magnitude of the exposure can be determined. DAC-h can be used to determine an individual’s dose when necessary. Air monitoring provides early warning of an immediate and significant exposure hazard and provides indications of the need for special bioassay monitoring. The monitoring criteria contained in 10 CFR 835.402(c) do not establish required levels of detection capability, that is, the minimum detectable dose. For example, it may not be feasible to actually confirm intakes that will result in 100-mrem E50 particularly for bioassay measurements of some alpha emitting radionuclides. Therefore, monitoring thresholds should not be considered requirements on the sensitivity of a particular measurement. Furthermore, workplace monitoring and occupancy factors should be considered, as appropriate, in evaluating potential exposures and monitoring needs.

10 CFR 835.402(d) requires that “internal dose evaluation programs” be capable of demonstrating compliance with the dose limits stated in 10 CFR 835.202 (e.g., 5 rem committed effective dose in a year, or 50 rem committed equivalent dose to an organ or tissue other than the eye). In light of this requirement, there are three distinct situations for internal dosimetry programs: DOE-STD-1121-2008 31 1. Adequate technology. In this situation, routine bioassay measurements can show not only compliance with 10 CFR 835.202, but can be used to assess doses when E50 ≤ 100 mrem (the investigation level). An example of an “adequate technology” situation, that is, where there is no technology shortfall, is a routine urinalysis program for 3 H or a routine in vivo counting program for 137Cs: in each case, the MDA is less than the DIL. 2. Technology shortfall for routine bioassay. In this situation, the DIL is less than the MDA for practical routine bioassay, but special bioassay, triggered by workplace indicators, is available on short notice that can be used to show compliance with 10 CFR 835.402(d). An example of a “technology shortfall for routine bioassay” situation is a state-of-the-art internal dosimetry program for plutonium supplemented by vigorous workplace monitoring and controls. 3. No practical bioassay. In this situation, no bioassay method is available for the radionuclides in question, and no bioassay program, either routine or special, can show compliance with 10 CFR 835.202. An example of a “no practical bioassay” situation is routine worker exposure to the short-lived decay products of radon and thoron, in which no bioassay program can demonstrate compliance with the limits. In the “no practical bioassay” case, the only recourse in showing compliance with 10 CFR 835.202 is using representative air monitoring, tracking worker exposure in DAC-h or working level months (WLM), and performing dose assessments on the basis of the air monitoring results. For the short-lived progeny of radon and thoron, worker stay times and measurements of potential alpha energy concentration (PAEC) can be converted to potential alpha energy exposure (PAEE) in WLM. Alternatively, worker stay times and radon concentration measurements, with knowledge or assumption of the equilibrium factor, can be converted to equilibrium equivalent DAC-h or to PAEE in WLM. Depending on the reason for DAC-h tracking, dose of record may not be required to be assigned from DAC-h tracking, even for those radionuclides where the missed dose (from bioassay) is greater than 100 millirem committed effective dose. Per 10 CFR 835.402(d), the internal dose monitoring program needs to be able to demonstrate compliance with the dose limits. Accordingly, if use of available bioassay cannot demonstrate compliance with the dose limits or the purpose of the DAC-h tracking is to assess the dose (per 10 CFR 835.209(b)), then DAC-h tracking results would be used for the dose of record. In addition, there are routine situations where there is a technology shortfall (i.e., section 4.1, discusses technology shortfall as routine bioassay not capable of detecting doses of 100 millirem). In many of these situations, workplace indicators, such as tracking of exposures to derived air concentrations (DAC-h), do not trigger special bioassay evaluation yet they indicate that internal doses of 100 millirem or greater in a year are likely. For these situations, it is highly recommended that use of air concentration data, which are representative of the air the worker breathed, be used for assessing internal dose. This approach is preferable to use of bioassay results which indicate no detectable activity. For example, air concentration values indicating a 40 DAC-h exposure (100 millirem) should be considered for assessing internal dose if subsequent negative bioassay results are obtained based on an analytical process that is only capable of detecting exposures in excess of 100 millirem.

Initiating timely, special bioassay (e.g., fecal bioassay shortly after working in a plutonium environment with a respirator or in response to elevated workplace monitoring results) may be helpful in the detection of internal exposures with bioassay. A timely special bioassay may provide a better basis for determining internal exposures than DAC-h tracking results. DOE-STD-1121-2008 32 The basis for either assessing dose from DAC-h tracking, or not, when there is a bioassay program technology shortfall shall be documented in the site’s Internal Dosimetry Technical Basis Document. Use of air sampling to replace bioassay to calculate doses has a number of caveats. Most importantly, the air sample must be representative of the breathing zone air. A discussion of this matter is found in NUREG-1400, Air Sampling in the Workplace (September 1993). 4.2 REFERENCE LEVELS AND DERIVED REFERENCE LEVELS A reference level is a predetermined value of a quantity that triggers a specified course of action when exceeded or expected to be exceeded. Reference levels may be dose-based or intake-based. Derived reference levels are the measurement values for particular bioassay or air sampling results that correspond to a more general reference level under specifically defined circumstances. Some suggested reference levels are described below: • Verification Level, LV - The level of unexpected intake or dose at or above which an attempt should be made to determine if the intake is real. For example, this is the level at which special follow-up measurements should be obtained to confirm a high routine result. Below this level, it may be assumed, routine results are valid and default assumptions can be used to calculate and assign intake and dose. • Investigation Level, IL - The level of intake or dose (specified in the IDG as 100 mrem) at or above which a bioassay or air monitoring result should be investigated. The intent of this level is to investigate the circumstances and, to the extent reasonable, to determine actual conditions and parameters for dose evaluation, rather than use default assumptions. An investigation may involve special measurements, work history review, determination of material form, and modification of biokinetic parameters, and may culminate in a dose assessment. • Medical Referral Level, LMR - The level of intake or dose at or above which the medical staff shall be notified. The notification should be made as promptly as possible, but does not necessarily constitute an identified need for therapy. Some suggested numerical values for these levels are shown in Table II. Additional discussion about the investigation level and derived investigation levels is provided in the following sections. This discussion is warranted by the definition of a 100-mrem investigation level in the IDG. 4.3 INVESTIGATION LEVEL AND DERIVED INVESTIGATION LEVEL In the IDG, the investigation level is an E50 of 0.1 rem (0.001 Sv) from intakes occurring in a year for general employees. Special ILs for minors, visitors (i.e., members of the public entering a controlled area), and the embryo/fetus of a declared pregnant worker should not exceed 50 mrem (0.5 mSv). Throughout this document, IL refers to the IL for the appropriate group unless otherwise specified.

In cases where it is practical, feasible, and affordable, internal dose evaluation programs should have a goal of assessing intakes of radioactive materials that occur in a year and that deliver an E50 at the IL, that is, intakes of 0.02 stochastic annual limit on intake (SALI) for general employees and 0.01 SALI (or less) for declared pregnant workers, minors, and visitors. DOE-STD-1121-2008 33 With the exception of the IL, which is specified in the IDG on the basis of monitoring thresholds in 10 CFR 835.402, DOE sites are encouraged to consider using the alternative reference level quantities given in Table II. Table II. Example Reference Level Magnitudes Reference Levels (Amounts of Intake, Except for DOE IL) General Employee, Except Declared Pregnant Worker Minor, Visitor, Declared Pregnant Worker Intake (SALI) Corresponding E50 (rem) Intake (SALI) Corresponding E50 (rem)1 Verification Level, LV 0.02 0.1 0.005 0.025 DOE Investigation Level, IL 0.02 0.1 0.01 0.05 Medical Referral Level, LMR 1 5 1 5 1Note that in the case of a declared pregnant worker, the dose to the embryo/fetus is the dose to be considered, not the dose to the worker. 4.4 DERIVED INVESTIGATION LEVELS Derived investigation levels (DILs) are derived reference levels of routine individual monitoring results. Examples of DILs are bioassay results, such as organ or body contents, or excreta concentrations or excretion rates that indicate an intake resulting in a dose exceeding an IL. Other examples of DILs are workplace exposures, in stochastic DAC-h modified by a safety factor, which could lead to an E50 greater than an IL. Internal dosimetry programs shall establish DILs for each individual monitoring method applied for the analysis of all radionuclides to which workers are likely to be exposed and document the derivation of such DILs in the internal dosimetry technical basis documentation. The physical and chemical characteristics of the radioactive material which may be taken into the body should be taken into account in establishing DILs. If an internal dosimetry program chooses to use Reference Man (ICRP Publications 23 and 30) default parameters in conjunction with the biokinetic modeling recommended in ICRP Publications 30 and 54 in deriving a DIL, these choices shall be identified in the internal dosimetry technical basis documentation. If one radionuclide is used as a tracer for a mixture of radionuclides, the DIL shall be based on the dose from the entire mixture, not just the tracer radionuclide. 4.4.1 Factors Affecting the DIL for Bioassay Factors such as significant clearance of a radionuclide in less than a year (e.g., tritium), the frequency of bioassay monitoring, and the likelihood of multiple exposures during a year (or under chronic intake conditions) should be considered in establishing a DIL. The DIL shall be established so that a committed effective dose of one IL from all intakes in a year is likely to be detected by the monitoring program, i.e., the minimum detectable dose should be less than one IL. If a nonroutine or an unexpected intake of a radionuclide or group of radionuclides occurs, the minimum detectable dose may be calculated assuming a single intake that occurred on the date of the intake, if known, or the date that would result in the largest committed effective dose. See Section 7.3.2.2 for guidance on assessing dose when time of intake is not known. If intermittent or chronic intakes are expected, the minimum detectable dose shall be calculated assuming a chronic intake during the sample period.

For nonroutine or unexpected intakes, the DIL for each independent radionuclide or group of radionuclides ensures that a committed effective dose of not more than one IL would be missed in the year from intakes of that radionuclide or group. DOE-STD-1121-2008 34 If it is known or is likely that an individual has or could have intakes during the year from different sources that could result in doses above the IL, appropriately smaller DILs shall be determined and the basis for those DILs included in the internal dosimetry technical basis documentation. 4.4.2 Calculating the Derived Investigation Level for a Given Sample Frequency The IDG states that an IL of 100 mrem (0.001 Sv) of committed effective dose from all intakes occurring within a dosimetric calendar year shall be used to establish DILs, and thus put an upper limit on the MDA for measurements. The desired value of the MDA may be further reduced by the need to confirm intakes by special follow-up bioassay: for rapidly clearing nuclides, a follow-up urine sample will generally contain a lower concentration of analyte than the initial unexpectedly high sample, but this lower concentration must still be detectable. There are at least two approaches to calculating DILs as a function of sampling frequency. One acceptable alternative is to set a derived screening level based on an intake corresponding to some fraction of the IL (e.g., E50 = 1/10 IL or 10 mrem for workers). The intent is to ensure that the reason and conditions of the intake are understood and that multiple intakes whose total would lead to an E50 approaching the IL could not be missed. This derived screening level is for each intake, while the IL is for all intakes in a year. This simple approach is acceptable for exposures to multiple independent sources and is adequate for use by DOE sites. A second acceptable alternative is to compute a DIL as a function of sampling frequency. With a sampling frequency of f samples per year (e.g., f = 12 per year for monthly samples), the goal of being able to detect 100 mrem of E50 from all intakes in a year means that each analysis must be capable of detecting 100 mrem ÷ f. Thus, a yearly investigation level of 100 mrem results in a sample investigation level (ILs) of (100 mrem/year)/(f samples/year). For example, for f = 12 per year, ILs = 100/12 = 8.3 mrem per sample. Thus, there is a detection-level penalty for frequent sampling. The latter approach is especially important for radionuclides with short physical or biological half-lives such that multiple sampling in a year is essential. The screening level approach described above provides relief from complicated calculations by establishing the screening level per intake, below which a bioassay result can be disregarded, regardless of sampling frequency. The sample investigation level is used to compute the DIL. Let IRFu(t) denote the intake retention function for urinary excretion at time t following a single acute intake (Bq per day excreted in urine per Bq of intake) (see Potter, 2002 for IRFs). Let V u be the urine excretion rate for Reference Man, 1.4 liters per day. Let the effective dose conversion factor be denoted by e50 (i.e., the committed effective dose per unit of activity of the radionuclide taken in by a specified route in Sv per Bq) tabulated in ICRP Publication 68 (1994b). Let Cu(t) denote the observed concentration of analyte in urine at time t. Then the committed effective dose is

u 50 u u (50) ( ) (50). ( ) VE I e C t e IRF t = ⋅ = ⋅  (1) Rearranging the equation to solve for concentration, we have u u 50 u ( )( ) (50) IRF tC t E e V = ⋅ ⋅ ⋅  (2) This equation is used to determine the DIL(f) for a given sampling frequency f by setting E50 to DOE-STD-1121-2008 35 the ILs (=IL/f) and evaluating the IRFu(t) at t = (365 days per year) ÷ (f samples per year), that is, the longest period between a possible intake and bioassay: u u ( 365/ )( ) (50) IL IRF t fDIL f f e V ⋅ = = ⋅ ⋅  (3) To meet the performance objectives described in the IDG, the MDC or the MDA must be less than the DIL(f). Use of Equation (3) is shown in Example 4.1. Example 4.1. DIL for Type D Natural Uranium For the more complicated case of several independent sources of radionuclides or groups of radionuclides, a more elaborate method may be needed. In many cases, the number of independent sources to which a worker will be exposed in a year is not known until the end of the year. Nonetheless, one can identify the formalism needed to calculate DILs for many independent sources. The concept of acceptable minimum detectable dose (AMDD) in the case of multiple independent sources is introduced as a tool to help calculate DILs. The AMDD is a dose value less than the IL by a factor that depends on the number and nature of independent sources a worker may be exposed to. To determine the AMDD in a year for a given radionuclide or group of radionuclides, j, it is necessary to consider the number of independent sources n to which an individual worker may be exposed, as shown in Figure 1. For each independent source, j, a judgment must be made concerning whether intakes of that group are characterized as “rare, single” intakes or whether there is a possibility of multiple or chronic intakes. In the latter case, if the nuclide is rapidly clearing, then a dummy variable, pj, is set to 1. Assume we have chosen a sampling frequency f = 12 samples per year. For Type F uranium, daily urine excretion (30 days) = 0.0018 Bq per day per Bq of intake (ICRP 1997). From p. 125 of ICRP Publication 78 (ICRP 1997), e(50) for inhalation of Type F 234U = 6.4E-7 Sv/Bq, 235U = 6.0E-7 Sv/Bq, and 238U = 5.8E-7 Sv/Bq. Natural uranium is a mixture of these three isotopes. Since 234U gives the highest dose per unit intake by a small margin, one may conservatively use the value for 234U. Then, the DIL becomes 𝐷𝐷𝐷𝐷𝐷𝐷 Inh Type F nat𝑈𝑈,  𝑓𝑓 = 12/year)   =    0.001 Sv y-1 ⋅ 0.0018 𝑑𝑑−1 12 𝑦𝑦−1 ⋅ 6.4 × 10−7 Sv Bq-1 ⋅ 1.4 𝐿𝐿 𝑑𝑑-1  =   0.167 Bq 𝐿𝐿−1×1 dpm/0.0167 Bq   =   10.0 dpm 𝐿𝐿−1 (total uranium 𝛼𝛼 activity) Since the intake retention fraction decreases as the interval between samples increases (i.e., as f decreases), and the sample frequency f appears explicitly in the denominator of the DIL equation above, there is some optimum choice of f that requires the least detection capability. However, since annual cost is directly proportional to f, there are trade-offs between cost and detection capability. DOE-STD-1121-2008 36 For “rare, single” intakes or for possible multiple or chronic intakes of slowly-clearing nuclides, pj = 0. The AMDD (mrem per year) for each independent source then becomes 1 if 0, if 0. j j j jn j j AMMD IL p or ILAMDD p p = = = = ≠ ∑ (4)

Figure 1. Alternative Logic Flow Chart for Determining the “Acceptable Minimum Detectable Dose” (AMDD) and DIL for Each Radionuclide or Group of Radionuclides When Exposure to Multiple, Independent Sources Is Possible DOE-STD-1121-2008 37 In other words, AMDDs for rare intake radionuclides and slowly clearing multiple or chronic intake radionuclides are equal to the IL, and those for possible multiple intake or chronic intake groups that clear quickly are reduced by a factor of 1/k, where k is the number of radionuclides or radionuclide groups j for which multiple or chronic intakes are possible. A lower limit on the DIL for radionuclide group j as a function of sampling frequency can be determined. This limit is the detection sensitivity needed for a bioassay measurement, that is, the minimum change one would need to detect in each bioassay measurement to detect a series of small intakes resulting in the AMDD for group j. This detection sensitivity or minimum change in amount, ∆Amin, is given by u min u ( 365/ ) ( ) Δ , (50) j j AMDD IRF t f DIL f A f e V ⋅ = ≥ = ⋅ ⋅  (5) where AMDDj is substituted for the IL, and the other terms are as defined above. This formalism accounts for the problem of multiple independent sources. The sampling frequency that makes minimum demands on analytical technology in terms of its detection sensitivity for analyte in bioassay samples is that frequency for which the ∆Amin is maximized. This sampling frequency can be found by setting min( ) 0d A df ∆ = ⋅ (6) Use of this equation can help determine the optimum sampling frequency for radionuclides for which the MDA is undesirably high. Example 4.2 shows this approach for tritium. DOE-STD-1121-2008 38 Example 4.2. Maximizing the Detection Sensitivity for Chronic Intakes of Tritium While use of the second method for establishing DILs may provide assurance that there is no possibility of missing intakes resulting in doses at or above the IL, it may be too complicated for practical implementation. It may be possible to apply the averaging techniques of Strom and McGuire (Strom and McGuire 1993) as detailed in NUREG 1400 (Hickey et al. 1993) to improve the counting statistics, and thus reduce the MDA for a given bioassay technique, but this has been established only for air monitoring. To illustrate the dependence of the detection sensitivity on f, consider the simple IRFu(t) for 3H: where k is a normalizing constant. Substituting 365/f for t and putting this in the ∆Amin equation, we have For the case of 3H, the sampling frequency that makes minimum demands on analytical technology is The solution to this is found by setting the term in parentheses to zero, giving The interval between the samples is simply the average clearance time τeff = 1/λeff = 14.4 days for 3H. A plot of the 3H∆Amin for a constant total annual missed dose as a function of sampling frequency is shown in Figure 2. If sampling is done more often than once every τeff, a lower ∆Amin (better analytical lab capability) is needed to see intakes resulting in the AMDD. DOE-STD-1121-2008 39 Figure 2. Plot of the Normalized Detection Sensitivity as a Function of Number of Samples per Year for 3H 4.4.3 Factors Affecting the DIL for Air Sampling A given air monitoring result may indicate a concentration higher or lower than that in the air

breathed by a particular worker or workers. How well an air sample reflects the concentration actually inhaled by a worker is called “representativeness.” Bioassay results, which are specific to an individual, do not have this property. Efforts to correlate bioassay measurements with workplace air monitoring have shown that intakes predicted on the basis of general area (GA) air monitoring results may have limited correlation with intakes based on bioassay results. Breathing zone (BZ) air samples are more representative. Air monitoring results, depending on where the sampler input is located, may underestimate intakes due to the “Pig Pen” effect2, in which air is more contaminated near a worker than at some distance away. The explanation for the Pig Pen effect is simply that the worker is generating the aerosol. It is important because it impacts the degree to which an air sample represents the concentration breathed by a worker, and it leads to the need to consider a safety factor when formulating a DIL for air monitoring. For an IL of 100 mrem of E50, s40 DAC h , ( ) DIL Safety Factor to allow for poor representativeness − = (7) where the subscript “s” denotes “stochastic.” Depending on the location of the air sampler with respect to the worker’s breathing zone, the value of Safety Factor may be in the range of 1 to 10, based on NUREG-1400 (Hickey et al. 1993) and Caldwell’s work (Caldwell 1972). Caldwell showed that, for plutonium work, fixed station air samplers tended to dramatically underestimate intakes assessed from fecal samples, and that lapel-type breathing zone air samplers more accurately corresponded to intakes 2Named after the Charles Schultz character in the Peanuts™ comic strip who walks around in a cloud of dust and debris, this term is attributed to the late Roger D. Caldwell. 0 1 0 50 100 150 200 250 300 350 400 Frequency, N (samples per year) N or m al iz ed D et ec tio n Se ns itiv ity DOE-STD-1121-2008 40 predicted using the 1966 ICRP lung model and fecal data. He also showed wide variability between breathing zone air results and general area air results, with median BZ/GA ratios between 3 and 8, and 90 percentile ratios from 9 to 26. 4.4.4 Supplementing Routine Bioassay Programs When DIL < MDA DOE’s 10 CFR 835.402(c) requires that, with a likelihood for 0.1 or more rem of E50, a worker must be on a dose evaluation program. There is no requirement that the program be able to detect 0.1 rem of committed effective dose, only that it has to detect 5 rem of committed effective dose, as in 10 CFR 835.402(d). To gain insight on the question of detection capability, one may examine requirements for external irradiation. There is the same 0.1-rem threshold for external monitoring, but an additional requirement that external dosimeters be accredited by the U.S. Department of Energy Laboratory Accreditation Program (DOELAP). Since 10 CFR 835 is a requirements document, then the standards in the DOELAP standards are requirements. Thus, personnel dosimeters must be able to detect 0.03 rem in several categories of radiation exposure. The practice for external irradiation is to require not only detection capability at 30% of the monitoring threshold, but also fairly precise and accurate detection capability at that level. By analogy, one might consider it desirable for an internal dosimetry program to be capable of detecting E50 values in the same range. However, this is not always practical or even feasible.

There is a technology shortfall for routine bioassay programs when the DIL is lower than the MDA. When a bioassay program has DIL < MDA, BZ or personal air monitoring may be implemented to supplement the routine bioassay program, as illustrated in Example 4.3. Personal air samplers are often more representative than fixed samplers. However, personal air samplers have a lower flow rate than most fixed air samplers. Example 4.4 shows how averaging of periodic results can be used to lower the MDC. 4.4.5 A Potential Technology Shortfall for Breathing Zone Air Sampling Breathing zone air sampling may not be adequate in facilities where 238Pu or another high specific activity alpha emitter is processed. High specific activity radionuclides usually have shorter half-lives than lower specific activity isotopes. The problem with high specific activity radionuclides arises from the fact that a small number of particles can be significant from a dose standpoint as illustrated in Example 4.5. However, as shown in Example 4.6, a similar concern does not exist for isotopes with lower specific activity, such as 239Pu. Thorough discussions of the problems with detecting and quantifying intakes using personal air samplers are given by Birchall et al. (1985, 1986, 1987, 1991) and by Scott et al. (1997). There is historical precedent for a BZ or personal air monitoring program supplemented by an aggressive fecal sampling program in NRC-licensed plutonium facilities. A facility operated in the 1960s and 1970s in Parks Township, Pennsylvania, by NUMEC, ARCO, and most recently by Babcock & Wilcox (Caldwell 1972), which processed reactor-grade plutonium, did not have significant trouble with the “countable number of particles” problem discussed in Example 4.5. DOE-STD-1121-2008 41 To illustrate the detection capability of breathing zone air monitoring, consider the DAC for Type S 239 Pu of 6E-11 µCi/mL (10 CFR 835, Appendix A). Multiplying by 2.4E9 mL/year breathed by Reference Man, one derives DOE’s deterministic or nonstochastic annual limit on intake (NALI) for Type S 239 Pu as 1.44 E-1 µCi = 144 nCi (5.33 kBq). The complementary “5-rem” (“0.05 Sv”) stochastic annual limit on intake (SALI) calculated from the effective dose per unit intake coefficients, e(50) in ICRP 68 (in SI units, as used in ICRP Publications) is as follows: Converting to conventional units, For Type S Pu, the SALI is 162 nCi (6000 Bq). Then, 2% of a SALI (that is, the intake that would result in a E50 of 100 mrem) is 3 nCi (120 Bq), or 7200 dpm of Pu. (The SALI for Type M material is about 1/4 of the SALI for Type S.) Suppose a worker was exposed to an atmosphere in which, breathing at Reference Man’s rate of 20 liters per minute, he would experience an intake of 2% of a SALI. A BZ or personal air sampler operating at 20 L/min would collect this same 7200 dpm of Pu activity. A lapel air sampler operating at 1.8 L/min would accumulate about 640 dpm (11 Bq). Thus, for a single air sample, there is no difficulty (in the sense of counting statistics problems) achieving detection capabilities comparable to those that the DOE requires for external radiation monitoring using BZ or personal air samples for a single filter. Personal air sampler filters are likely to be changed every day, or 250 times in a year. Thus, the 7200 dpm, which is 2% of the SALI, could be on one filter or spread among many or all. The minimum detectable intake for uniform, chronic exposure based on 250 samples is higher than the minimum detectable intake for a single, acute exposure. See Example 4.4.

DOE-STD-1121-2008 42 Example 4.3. Use of Breathing Zone Air Samples to Supplement Routine Bioassay for Plutonium Another benefit of BZ air monitoring programs is that they give workers feedback about work practices. The experience at Apollo, Pennsylvania, showed that workers develop better radiological control habits based on BZ air sample results. It is well known that bioassay is much more accurate than BZ or personal air monitoring when bioassay results are available and adequate. However, when bioassay methods are not adequate or unavailable, BZ or personal air monitoring shall be used. (See 10 CFR 835.209(b).) When there is technology shortfall for routine bioassay, DOE sites should consider using BZ or personal air monitoring programs to supplement their routine bioassay programs. Such use should be tempered with an understanding of the limitations described in Example 4.5. Example 4.4. Improving Detection Capabilities of Air Sampling Using Averaging The minimum detectable average concentration for repeated BZ or personal air samples over a year or other period of time can be reduced by averaging the original raw data, as described in the Appendix to NUREG-1400 (Hickey et al. 1993a, Strom 1993). The simplest case is when independent activity measurements are made of a sequence of samples for which large numbers of counts (i.e., more than 50) are collected and for which the following remain identical between samples: background count times and rates, sample count times, and counting yields. In such a case, the MDA for the sum of n samples is larger than that for a single sample: MDA(n) = √ .MDA(1). Conversely, the minimum detectable average concentration (MDC) for n samples is smaller than the minimum detectable concentration (MDC) for a single sample: MDC(n) = MDC(1)/ √ , when sample volumes or masses are all equal, equal sample collection times are used, and collection efficiencies are equal. Although the MDA for such pooled samples increases by √ , the volume or mass in which this activity is found increases by a factor of n, resulting in a net decrease in MDC by a factor of √ /n = 1/ √ . In general, samples may have varying count times, background count rates, counting efficiencies, collection efficiencies, and collection times. Exact time-weighted formulas for MDC and decision level (DL) are given for the general case in the Appendix to NUREG-1400, and exact formulas are provided for both large and small numbers of background counts (Hickey et al. 1993a). This methodology is useful in situations where daily, weekly, or monthly concentration measurements must be compared to an annual limit. It is also useful in determining the detection capabilities of a measurement program. This work shows the importance of reporting measurements and their standard deviations as observed, rather than “censoring” them by reporting them as “less than” values. An alternative to averaging is to physically combine air filters containing long-lived material. For example, if a worker had 200 separate personal air sample filters during a year, they could be combined and the composite analyzed as a single sample. If the material were a penetrating photon- emitter, the ensemble of filters could be counted directly by gamma spectroscopy. If the material were an alpha-emitter, radiochemistry would be necessary. DOE-STD-1121-2008 43 4.4.6 Performance Specifications for a Bioassay Laboratory 10 CFR 835.402 requires internal dose monitoring programs implemented to demonstrate

compliance with 10 CFR 835 be accredited by the U.S. Department of Energy Laboratory Accreditation Program (DOELAP) for Radiobioassay (DOE 2015). Radiobioassay laboratories utilized by the internal dose monitoring programs will be evaluated against the requirements of the “Department of Energy Laboratory Accreditation Program for Radiobioassay” (DOE 2019a) which incorporates the recommendations of ANSI/HPS N13.30-2011, “Performance Criteria for Radiobioassay” (HPS 2011c). In addition, they may wish to consider the requirements of ANSI N42.23-1996, “Measurement and Associated Instrumentation Quality Assurance for Radioassay Laboratories” (ANSI 1996). Additional specifications for the bioassay or service laboratory should be negotiated between the site and the laboratory. Example 4.7 gives performance specifications for a radiobioassay laboratory. DOE-STD-1121-2008 44 Example 4.5. Potential Technology Shortfall for Breathing Zone Air Sampling of High Specific Activity Alpha Emitting Nuclides For high specific activity alpha emitters, a single large particle on an air sampler filter may give erroneous results, a phenomenon that can be described as the “countable number of particles problem.” In facilities where 238Pu is processed, it may be difficult to use BZ or personal air monitoring to control intakes near the level of 2% of a NALI. Using the methods in ICRP Publication 66, and a density of 11 g/cm3 for plutonium oxide (ICRP 1994a and p. 1.7 of Faust et al. 1988), the table below was calculated. The equivalent physical diameter also accounts for slip correction and thermodynamic effects, both important at small particle sizes. The table shows that one particle with an aerodynamic diameter of 5 μm is approximately 2% of a NALI. With Monte Carlo analysis, Scott et al. (1997) show that calculated average intake of high specific activity alpha emitters, in DAC-h, is not an operationally useful quantity. They used a light activity breathing rate of 1.5 m3h-1, a density of 10.0 g cm-3, an AMAD of 5 μm, and a GSD of 2.5 and calculated the intakes of 10,000 workers exposed. In an 8 DAC-h exposure, 9,831 had no intake, 4 had intakes greater than one ALI (that is, 2,000 DAC-h or 600 Bq of 238Pu), and 165 had intakes ranging from a fraction of a DAC-h to nearly 2,000. All intakes resulted from inhaling a single particle of 239PuO2. Thus, the average intake computed for the group of workers, would both overestimate the intakes of the vast majority of individuals and seriously underestimate intakes of the more highly exposed individuals. DOE-STD-1121-2008 45 Example 4.6. The Number of Particles for Breathing Zone Air Sampling of a Lower Specific Activity Radionuclide In the previous example, it was shown that breathing zone air sampling in a facility that handled 238Pu might not be useful. For lower specific activity material, there is no similar problem. For example, “6%” plutonium aged 14.4 years, has 50% ingrowth of 241Am (Rittmann 1993), a specific activity of 3.44E9 Bq/g of α-emitters, and an α-NALI of 458 Bq (bone surfaces). Using the same density for plutonium oxide and calculational approach as in Example 4.5, gives the table below. It shows that the number of particles corresponding to 2% of a NALI does not create a problem until the AMAD > 10 μm. It is important, however, to minimize accidental filter contamination by even one “large” particle. One 10-μm particle corresponds to an HT=bone surfaces,50 of about 60 mrem and an E50 of about 5 mrem.

DOE-STD-1121-2008 46 Example 4.7. Example of Performance Specifications for a Bioassay Laboratory The radiobioassay laboratory shall meet the contractual minimum detectable amounts as listed in [DOE site to provide specific list]. Control sample results shall, as a minimum, meet the criteria concerning relative bias statistics within -0.25 to +0.50 and the relative precision statistic shall be less than or equal to 0.4. At the levels to be used in spikes, the bias and precision should normally be smaller than the limits in ANSI/HPS N13.30-2011. The radiobioassay laboratory shall verify that these limits are met. The radiobioassay laboratory (if required by 10 CFR 835) shall participate in the DOELAP for Radiobioassay. The radiobioassay laboratory shall achieve satisfactory results for all appropriate test categories. In addition, the radiobioassay laboratory should participate in traceability-testing for bioassay sample matrices offered through NIST’s Radiochemistry Intercomparison Program (NRIP). (Note: non bioassay matrices are not good indicators of bioassay laboratory performance.) The radiobioassay laboratory shall furnish the DOE site with all intercomparison data annually and/or upon request. The radiobioassay laboratory shall furnish the DOE site with all internal quality assurance and quality control (QA/QC) data upon request. The radiobioassay laboratory’s quality assurance program shall be implemented through an established documented plan. The QA program must also satisfy ANSI/HPS N13.30-2011. The radiobioassay laboratory will prepare and analyze reagent blanks and spiked urine and fecal samples for internal quality control. The number of QC spiked samples shall be at least 5% of the total samples analyzed and a reagent blank shall be analyzed with each set of samples. The reagent blanks will be used by the radiobioassay laboratory and the DOE Site, during audits and review of bioassay reports, to verify that all detection levels comply with the Contractual Detection Levels specified above. (The correct equation for verification of detection level is documented in HPS 13.30-1996.) The radiobioassay laboratory must satisfy initially and on a continuing basis certain quality control factors specified below concerning yields, resolution, contamination and control standards or a “stop work order” may be enforced until the problem(s) is resolved. The radiobioassay laboratory will report internal quality control results to the DOE Site Procurement Manager when requested. The DOE site may send, from time to time, blind spiked and/or blank samples to the radiobioassay laboratory. These sample results will be compared to the in vitro performance criteria documented in ANSI/HPS N13.30-2011 and will be used in conjunction with the radiobioassay laboratory’s in-house quality control results to determine if the radiobioassay laboratory is meeting the Contractual Detection Levels. (Note: A limited number of blanks are not a good indicator of the true MDA. It is better to use the lab’s QC results). (continued) DOE-STD-1121-2008 47 Example 4.7 (continued) Example of Performance Specifications for a Bioassay Laboratory Other factors that should be negotiated and put into the statement of work include turnaround time(s) for analytical results, especially for special bioassay; the need for prompt notification of results that exceed certain levels; and length of storage time for unused portion of samples or final analyzed preparation of samples (e.g., counting planchet) to allow for reanalysis or recounting of samples, if necessary.

The radiobioassay laboratory is required to maintain a QA manual that outlines responsibilities and also provides requirements for data control, document control, maintenance/test equipment calibration and checks, procedures, training, corrective action in the event of noncompliance, and traceability to standardizing bodies such as the National Institute of Standards and Technology (NIST) (when available). All instruments used for the analysis of the radionuclides in the bioassay program shall be properly “response”-checked before being used to analyze the DOE site’s samples. The results of the response checks shall be documented for each instrument that requires calibration (e.g., radiation detectors, scales, balances, etc.). All radiation detection instruments used for analysis of the radionuclides in the bioassay program shall be calibrated at least annually using NIST-traceable standards when they are available. A NIST certificate for all standards (when available) shall be retained by the radiobioassay laboratory and shall be made available to the DOE site for review. Additional Quality Control Factors Yields: The average yields determined for plutonium and strontium separated from urine and feces shall be at least 50% without restrictions, and at least 25% if it is determined that contractual minimum detectable amounts can be met. For americium and uranium, the average yields shall be at least 40% and 20%, respectively. Resolution: The resolution of α-particle spectrum energy peaks shall be less than 100 keV full width at half maximum. Contamination: The results of the reagent blanks shall be at least low enough to allow meeting the minimum detectable amounts. Any trend or sudden change towards increase in activity in blanks or their standard deviations that may cause the contractual minimum detectable amounts to be exceeded should be investigated and the cause eliminated. Contamination of the final fraction of one element with the nuclide of another element becomes important in alpha-particle spectrometry, particularly when it involves nuclides with alpha energies that cannot be resolved (energy peaks within one full width at half maximum of each other). Whenever potentially interfering foreign nuclides appear in the final fraction of any element, the cause for the contaminations should be identified and eliminated. If the magnitude of the contamination adversely affects the result, work shall be stopped until the problem is solved. However, work stoppage is not warranted for an isolated suspected contamination event. (continued) DOE-STD-1121-2008 48 4.5 MEASUREMENTS OF WORKPLACE RADON AND THORON CONCENTRATIONS, POTENTIAL ALPHA ENERGY CONCENTRATIONS, AND MEASUREMENTS OF (OR ASSUMPTIONS ABOUT) EQUILIBRIUM FACTORS 4.5.1 Measurements There are two objectives of radon/radon progeny monitoring and hence two sets of standards for these measurements. The two monitoring objectives are 1) to characterize in real time the concentrations that workers might be exposed to while in an area and 2) to establish the exposure of record that each worker actually receives. In the Air Monitoring chapter of the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a), these two types of monitoring are respectively referred to as air monitoring and air sampling. It will generally be found that meeting both objectives is best achieved using two different types of instruments.

Instruments used for both purposes shall measure either airborne radon or radon progeny concentration. If materials containing thorium-232 or its progeny are known to be present in the area, the instruments should also be capable of measuring airborne thoron progeny concentrations. Instruments used for air monitoring shall be real-time monitors that continuously measure and display results for periods of one hour or less. They should be placed to measure the highest concentrations to which workers in the area are likely to be exposed. Instruments used for air sampling shall be continuous instruments that make either time averaged or real-time measurements. They should be placed so as to measure as nearly as is practicable the concentrations to which workers are exposed. In areas with large gradients of concentration or equilibrium (e.g., outdoors), individual personnel monitors shall be used for each worker. Several good references are available for radon and thoron measurements (NEA 1985; NCRP 1990; Fortmann 1994). Sheets gives a recent review of indoor thoron with many references (Sheets 1993). 4.5.2 Equilibrium Factors If radon measuring instruments are used, radon progeny concentration should be inferred by application of an appropriate equilibrium factor. In general, equilibrium factors should be measured under a representative set of circumstances and for a representative time frame. If it is not practical to measure equilibrium factors, a default 222Rn equilibrium factor of 0.4 Example 4.7 (continued) Example of Performance Specifications for a Bioassay Laboratory Quality Control Spikes: ANSI/HPS N13.30-2011 requires that control sample results, as a minimum, have a relative bias statistics within -0.25 to +0.50 and a relative precision statistic of less than or equal to 0.4. At the levels to be used in spikes, the bias and precision should normally be smaller than the limits in ANSI/HPS N13.30-2011. The radiobioassay laboratory shall verify that these limits are met. DOE-STD-1121-2008 49 (ICRP 1993a; UNSCEAR 1993) may be used for indoor areas with normal ventilation rates and outdoor areas with radon sources no closer than 400 m (≈ 1/4 mile; Table III). Average indoor equilibrium factors increase with increasing particle concentration in air, and decrease with increased air exchange rate (James et al. 1988; James 1994; National Research Council 1991; NEA 1985; UNSCEAR 1993). For outdoor areas with local sources of radon and highly ventilated indoor areas, the appropriate equilibrium factor should be determined by concurrent radon and radon progeny measurements made over a set of conditions that present the range of equilibrium factors to be encountered when workers are present. These measurements and the rationale for their application to inferring radon progeny concentration shall be documented in the internal dosimetry technical basis documentation. Table III. Acceptable Default Equilibrium Factors for Radon (FRn) Location/Environment Default Equilibrium Factor (FRn) Indoors, normal ventilation 0.4 Indoors, unusual ventilation Measure Outdoors - no local radon sources 0.4 Outdoors with “local” radon sources Measure Appendix A contains a review of measurements of radon progeny equilibrium factors on which Table III is based. Appendix A also contains a brief review of published values of thoron progeny equilibrium factors. 4.5.3 Performance Criteria for Instruments Used at DOE Sites to Characterize Airborne Radon and Thoron and Their Progeny

The American National Standards Institute provides performance specifications for instruments for the measurement of radon, radon progeny, and thoron progeny in air (ANSI 1994a, 1994b). All instruments shall be operated using standardized approved operating procedures. All operators should be trained on these procedures prior to performing field measurements. Air Monitoring Instruments used for air monitoring shall have the following characteristics: • a response rate that is limited only by radon progeny ingrowth (i.e., a full-scale response time of about 4 hours; does not apply to thoron), • a sensitivity to environmental influences that complies with the applicable parts of ANSI N42.17A-2003 and ANSI N42.17B-1989, • a coefficient of variation of no more than 15% when making one-hour measurements of constant, normal background concentrations, and • a calibration bias of no more than 10%. Enclosing the instrument in a protective housing may be necessary to limit environmental influences. DOE-STD-1121-2008 50 Air Sampling Instruments used for air sampling shall have the following characteristics: • a sensitivity to environmental influences that complies with the applicable parts of ANSI N42.17A-2003 and ANSI N42.17B-1989, • a coefficient of variation of no more than 15% when making 170-hour measurements of constant, normal background concentrations, and • a calibration bias of no more than 10%. 4.5.4 Participation by DOE Sites in an Intercomparison Program for Radon Instrument Calibration, Precision, and Accuracy Compliance of the measuring system(s) with the above performance specifications shall be demonstrated by subjecting a representative sample of instruments to periodic (annual if possible) radon and/or radon progeny comparison exercises, if and when such exercises are conducted by DOE laboratories. If and when the DOE Laboratory Accreditation Program (DOELAP) offers a radon measurements program, DOE sites shall participate in the DOELAP. 4.5.5 Calibration and Quality Control for Radon, Thoron, and Progeny Instrumentation All instruments shall be recalibrated at least annually. The lack of stability of some instruments may require that they be calibrated more frequently. Calibrations should be performed in a controlled atmosphere which is monitored with instruments whose flow rate and detection efficiency have been determined by reference to standards traceable to the National Institute of Standards and Technology, if such standards are available. Periodic functional tests should be performed at a frequency dependent on the performance history of the instrument. As a minimum, these tests will include checks of the airflow rate and detector efficiency. Replicate pairs of measurements should also be performed on a rotating schedule that covers all instruments at least once every two months. 4.5.6 Use of Engineering Controls for Management of Exposures to Radon, Thoron, and Their Short-Lived Decay Products The use of engineering control methods for radon and thoron should be based on cost-benefit analyses because they can be expensive to implement. Engineering controls for new building construction may be significantly cheaper than for existing construction. Guidance and model standards are available from the U.S. Environmental Protection Agency for reducing radon levels in existing construction (EPA 1989, 1991a, 1993, 1994b, 1994c, 2013). Such methods may be appropriate when the radon is due to DOE “activities” as defined in 10 CFR 835. Engineering controls for contaminated sites with elevated radon levels due to DOE activities may not be cost-effective, and personnel protective equipment or other radiation protection measures such as limiting stay times, performing work at times of the day when radon progeny levels are lower, etc., may be needed.

All new construction at DOE facilities that will be occupied for significant periods of time should be “radon-resistant” construction. References for radon-resistant construction methods are available from the U.S. Environmental Protection Agency and ASTM (EPA 1991b, 1994a; ASTM 2008). Making new DOE-STD-1121-2008 51 structures radon-resistant generally adds little to the cost of construction. DOE-STD-1121-2008 52 5 INDIVIDUAL MONITORING FOR INTERNAL DOSIMETRY 5.1 SCOPE OF PARTICIPATION IN INDIVIDUAL MONITORING PROGRAMS FOR INTERNAL DOSIMETRY Workers considered likely to have intakes resulting in an E50 in excess of 100-mrem are required by 10 CFR 835.402(c) to participate in an “internal dosimetry program.” Measurements from individual monitoring programs are needed as input to an internal dosimetry program. In the context of internal dosimetry, individual monitoring includes routine bioassay (mentioned in 10 CFR 835.402(c)) and/or personal air sampling (not mentioned in 10 CFR 835.402(c)). This section gives criteria for participation in individual monitoring programs, which include baseline, routine, special, and termination or task- ending bioassay and personal air sampling programs. Most radiation protection programs should be capable of preventing intakes through rigorous application of engineering and administrative controls. Under such controls, a good argument can be made that no one is likely to have an intake resulting in a E50 of 100 mrem. This may reduce the need for participation in a routine bioassay program (meaning scheduled periodic measurements) but does not eliminate the need for confirmatory or special bioassay monitoring. Likewise, the need for an internal dosimetry program is linked more to the potential for intake than the likelihood of intake. An internal dosimetry program must be available if sufficient quantities of radionuclides are present or handled at a facility in which accidental intakes resulting in 100-mrem E50 cannot be ruled out. 5.2 BASELINE INDIVIDUAL MONITORING: BIOASSAY Baseline monitoring involves determining appropriate baseline values for the worker at the start of employment or potential exposure. Internal dosimetry programs that must, out of necessity, be based on air sampling have no analog for baseline bioassay monitoring. The concept of establishing a baseline does not necessarily mean that baseline bioassay measurements be obtained. Administrative review of the worker’s history can lead to the conclusion that baseline measurements are not needed because the expected results are readily predictable (e.g., no detectable activity). Such a review can constitute acceptable baseline monitoring. If baseline measurements are needed, they should be completed before performing work requiring routine bioassay. Baseline measurements are appropriate for any of the following circumstances: 1) the worker has had prior exposure to the pertinent radionuclides and the effective retention in the body might exceed the screening level, 2) the exposure history is missing or inconclusive, or 3) the worker will be working with radioactive material which may be potentially detectable in bioassay due to non- occupational sources. Illustrations of baseline bioassay scenarios are given in Example 5.1. DOE-STD-1121-2008 53 Example 5.1. Baseline Bioassay Scenarios 5.3 PARTICIPATION IN ROUTINE INDIVIDUAL MONITORING PROGRAMS: BIOASSAY AND/OR PERSONAL AIR SAMPLING

Workers considered likely to receive intakes which could result in E50 values in excess of 100 mrem or who are at risk for such intakes shall participate in a routine individual monitoring program that includes bioassay and/or personal air sampling. Those workers are identified using criteria based on knowledge of the radionuclides, facilities, processes, and anticipated work. Criteria may be expressed in many forms, including quantity and form of material handled, type of work, or category of worker. There is no single method that is most cost-effective and technically correct for identifying those workers. Example 5.2 presents criteria for determining the need for routine participation in a bioassay and/or personal air sampling program and sample applications of those criteria. Example 5.3 gives instances in which personal monitoring is not needed. The ICRP (1988) recommends the order of preference for bioassay program data interpretation to be 1) direct in vivo measurement of body content, 2) excreta analysis, and 3) personal air sampling. However, the radionuclide or element being monitored and its characteristic radiations usually establish the type of monitoring performed. Participation in routine individual monitoring programs may be discontinued when sufficient facility history and experience is available to indicate that there is no need for a routine program. However, in such cases, a confirmatory monitoring program (see Section 5.7) may be of value. • A new employee in a plutonium facility would not require a baseline bioassay measurement if there was no prior potential occupational exposure to plutonium. However, a new employee at a plutonium facility who came from another facility where plutonium was a nuclide of concern should undergo baseline measurements if a termination bioassay was not performed by the former employer or work history information is absent. • Workers with potential exposure to uranium should receive baseline uranium urinalyses due to the ubiquitous and highly variable occurrence of uranium naturally and its possible presence in urine. • Workers with potential exposure to 137Cs should receive baseline whole body counts because of environmental 137Cs, present from worldwide atmospheric fallout, can be present in low levels in certain food products. DOE-STD-1121-2008 54 Example 5.2. Criteria for Participating in Individual Monitoring Programs Criterion 1: Quantity of radioactive material in process This criterion establishes a maximum working activity (MWA) or a mixture specific activity above which individual monitoring is recommended. The MWA is a quantity calculated using the nuclide stochastic ALI, and factors for such considerations as physical form of the material, containment or confinement methods, dispersibility based on the processing being performed on the material, occupancy, and a special form factor for DNA precursors. Examples of such formulations are provided in NUREG-1400 (Hickey et al. 1995) and ANSI/HPS N13.39 Appendix A (HPS 2011). Recent discussion among some health physicists suggests that the factors used in NUREG-1400 may be too liberal (i.e., too few people would be monitored), and this issue may be addressed in a future ANSI Standard. The mixture specific activity approach is described by Carbaugh et al. 2003a and applies to situations where radioactivity is essentially uniformly mixed with a large volume or mass of inert material (e.g., contaminated soil). Criterion 2: Worker training and tasks

Workers with Radiation Worker II training and who work with radioactive materials may be scheduled for routine bioassay and/or routine personal air sampling. This is a very broad-scope practice, giving rise to large programs. While it is easy to implement, it is likely to result in requiring personal internal dosimetry measurements of workers who are not likely to exceed 100 mrem of E50. The cost of the unnecessary measurements is a tradeoff for less scrutiny of actual worker assignments. Criterion 3: After-the-fact determination of bioassay need based on actual work performed (does not apply to air sampling) An aggressive program with continuing checks on worker potential exposures (e.g., entries into contamination areas or under specific radiological work permits) may be able to retroactively determine the need for bioassay based on actual work. Such a program might review a worker’s actual activities during the course of the last routine bioassay interval (e.g., one year) and determine that no potential for exposure occurred. Under such circumstances, the bioassay measurement which might otherwise be routinely obtained could be omitted. This practice calls for close review of an individual worker’s activities. The cost savings for omitted bioassay must be weighed against the cost of work history review to determine the net cost savings. Criterion 4: Use of respiratory protection to limit intake and dose When respiratory protection is used to limit intake of radioactive material, 10 CFR 835.403(a)(2) requires that air monitoring be done, as necessary, to characterize the hazard. In addition, this Technical Standard recommends that workers participate in routine bioassay monitoring if respiratory protection is used to limit the intake of radioactivity (i.e., respiratory protection factors are being used to limit the estimated intake of radioactivity). Routine bioassay may be omitted if respirators are used as a matter of conservative protocol without any actual indications of airborne radioactivity, or air sample results indicate that the worker would not have been at risk of exceeding the 100-mrem investigation level without respiratory protection. Workers who use positive pressure suits should undergo routine bioassay or have provisions for breathing zone air sampling within the suit. (continued) DOE-STD-1121-2008 55 Example 5.2. (Continued) Criterion 5: Long-term chronic exposure to air concentrations exceeding 2% of the DAC This condition can apply to facilities that have low-level airborne radioactivity but do not meet the criteria for posting as airborne radioactivity areas. Caution: just because an area does not require posting as an airborne radioactivity area does not mean that individual monitoring is not needed. Where routine air concentrations never exceed posting requirements but exceed 2% of the DAC, the need for individual monitoring should be based on potential stay times in those areas. Continuous (or significant) occupancy over a year would suggest individual monitoring is needed. Criterion 6: Short-term chronic airborne exposure, or multiple acute airborne exposures Criteria 6 and 7 may be particularly useful for addressing supervisory, walk-through, and inspection staff who do not actually handle or process radioactive material. The derived concentration threshold for individual monitoring (Cair, in terms of fractional DAC) can be calculated using an exposure fraction for the worker (fw), as follows:

Criterion 7: Tracking individual exposure in DAC-h Individual work assignments and concentrations are tracked to determine cumulative exposure in DAC-h. Once a worker exceeds 40 DAC-h, bioassay should be performed (if feasible). This method implies the use of a DAC-h tracking log. Such a log might be continued for a worker over the course of a year and then zeroed out at the start of a new year. One issue to be resolved by the facility is what to do with DAC-h if the total never exceeds 40. For this situation, sites should still provide intake and dose assessments based solely on DAC-h. 10 CFR 835.702(b) specifies thresholds for not needing to record internal dose for certain monitoring results. These include a monitoring result estimated to correspond to an individual receiving less than 10 mrem committed effective dose, or 4 DAC-h, not to exceed 100 mrem committed effective dose or 40 DAC-h over the year. DOE-STD-1121-2008 56 Example 5.3. Circumstances Not Requiring Individual Monitoring 5.3.1 Exposure Monitoring Thresholds for Radon and Thoron Progeny Since there is no practical bioassay for radon and thoron, exposure monitoring is required when individuals have the potential to be exposed in excess of the dose levels given in 10 CFR 835.402(c) requiring monitoring. It is important to emphasize that the radon and thoron exposure monitoring thresholds are exposure-based (WLM or DAC-h) versus concentration-based thresholds because of the dynamic nature of radon concentrations. The requirement in 10 CFR 835 is that monitoring be provided for workers who are likely to receive a potential alpha energy exposure (PAEE) above background that would lead to a committed effective dose E50 of 100 mrem in a year. The corresponding exposures are 0.2 WLM for 222Rn progeny and 0.6 WLM for 220Rn progeny. Monitoring would normally include breathing zone air sampling using lapel air samplers or etched-track detectors, or fixed air monitors with records of stay times. Because of compelling special circumstances, a few contractors have been able to get regulatory relief under 10 CFR 820.62. The problem arises from the inability at these sites to distinguish between natural and occupational sources of radon and thoron exposure. At these sites, monitoring is provided for workers who are likely to receive a PAEE including background that would lead to a committed effective dose E50 of 200 mrem in a year: this is 0.4 WLM for 222Rn progeny and 1.2 WLM for 220Rn progeny. Both approaches to a monitoring are based on exposure, which includes both air concentration and amount of time breathing the air. It is important to point out that workers may be permitted to work in significant concentrations of potential alpha energy for short periods of time with no personnel monitoring, providing they don’t exceed the likelihood of receiving an E50 of 100 mrem or 200 mrem (depending on whether the site has obtained an exemption from 10 CFR 835 for radon). 5.4 SPECIAL BIOASSAY PROGRAM Special bioassay shall be initiated when off-normal conditions occur or there are indications that an intake needing assessment may have occurred. Criteria for identifying those conditions typically can include personal contamination, high air sample results, uncontrolled spread of contamination, or expressed worker concerns. The response to these conditions is the performance of bioassay measurements outside the envelope of routinely performed baseline, scheduled periodic, and termination or ending work measurements. The reason for and interpretation of these special measurements should be clearly identified. The role of special measurements is to confirm or rule out the initial indication of an intake, to determine the radiological significance of confirmed intakes, to indicate the need for work restriction or dose reduction therapy, and to begin the dose assessment process.

Special bioassay measurements may include the same types of measurements as those performed • Radioactive materials are in a sealed source or special form. • Radioactive materials are packaged in accordance with Department of Transportation specifications. • Quantities of radioactive material in process are less than 2% of an ALI. DOE-STD-1121-2008 57 for routine monitoring (e.g., in vivo measurement, urinalysis) and may also include additional types of measurements (e.g., fecal analysis, wound counting). Some criteria for initiating special bioassay can be found in the RadCon Standard (DOE-STD- 1098-2017). Typically, site technical basis documents or programmatic manuals provide additional guidance. The criteria of Example 5.4 should be considered more as qualitative guidelines than quantitative requirements. The decision to select any particular contamination level as a criterion for initiating special bioassay is highly subjective. For example, a hot particle on a shoe cover would not necessarily warrant special bioassay, even though the contamination level may exceed the alpha or beta/gamma contamination level shown above. Likewise, a single spot of contamination on the side of the face would be less likely to warrant special bioassay than substantially lower levels of contamination covering the mouth and nose area. While it is certainly conservative to perform bioassay when any of the listed criteria are exceeded, an excellent internal dosimetry program will factor in the unique aspects of each occurrence and exercise good professional judgment in prescribing special bioassay. There are potential pitfalls in relying on some indicators as a basis for not performing special bioassay. For example, no detectable activity on nasal smears following a suspected inhalation does not necessarily mean that no intake occurred - a worker who has nasal congestion or is a mouth breather would not necessarily show activity detectable by nasal smears following an inhalation intake. Wounds involving alpha-emitting nuclides need special attention because the contamination could be completely shielded by overlying skin, tissue, blood, or serum moisture. Blood smears should be dried before counting with an alpha detector. In some cases, workplace detection methods can be adequate to moderate the need for immediate bioassay measurements. For example, high-energy beta or photon emitters such as 90Sr, 137Cs, and 60Co can be readily detected using portable Geiger-Müller (GM) survey meters. The typical sensitivity of such instruments is sufficient to determine the relative severity of a potential intake by a wound. If contamination is not detectable by these instruments at the time of the injury, then it is highly unlikely that there is any significant wound intake. This knowledge can permit a more relaxed approach to special bioassay, rather than precipitate a crisis response. Deciding the duration and extent of a special bioassay program also calls for professional judgment. It should be recognized that early excreta bioassay (collected earlier than 1 to 2 hours following the intake) will not necessarily reflect sufficient equilibrium to allow an accurate assessment of intake. Urine collected earlier than 1 hour after intake is likely to reflect the pre-intake condition. Likewise, feces voided within a few hours of an inhalation intake may be too early to have permitted passage of radioactivity through the gastrointestinal (GI) tract. In vivo measurements made shortly after intake may also reflect rapidly changing clearance. Residual external contamination on an in vivo bioassay subject is sometimes a problem near the time of intake. Thus, multiple bioassay measurements over several days following an intake provide a better tool for quantifying the magnitude than a single sample. These may include longer term measurements at weeks, months, and even years after an intake to accurately characterize the biokinetics and provide accurate intake and dose assessments.

DOE-STD-1121-2008 58 Example 5.4. Criteria for Commencing Special Bioassay 5.5 TERMINATION AND ENDING-TASK BIOASSAY PARTICIPATION When a worker completes an assignment requiring routine bioassay, an ending-task bioassay measurement is used to indicate the worker’s status when the potential for further exposure has ended. Ideally, this measurement should be made as soon as the work assignment is completed. If the measurement is not made until employment is ended, then the measurement is actually an employment termination measurement and documents the status of the worker when no further occupational exposure under that employer will occur. Ideally, the termination measurement would be performed on the last day of employment. The need for both ending-task and termination samples is a matter of company policy. If ending-task measurements are performed and the cognizant radiation protection organization is confident that no further potential for intake existed, then an employment termination bioassay is probably not needed. For practical purposes, the ending-task measurement may be considered the release of a worker from requirements for further bioassay. Special bioassay should be initiated if any of the following criteria are met (Fauth et al. 1996, Carbaugh 2009): • Nasal or mouth smears, nose blows, or sputum samples that indicate above background levels of radioactivity • Any contaminated wound • Contamination on protective clothing in excess of 10,000 dpm-alpha or 100,000 dpm- beta/ gamma per 100 cm2 if no respiratory protection is in use • Unplanned spread of contamination on accessible surfaces in excess of 1500 dpm- alpha or 15,000 dpm-beta/gamma per 100 cm2 if no respiratory protection is in use • Any detectable general facial contamination in excess of 200 dpm-alpha or 4,000 dpm-beta/ gamma per 100 cm2 • Detectable contamination on the skin, other than the facial area, in excess of 1000 dpm-alpha or 100,000 dpm-beta/gamma per 100 cm2 • Detectable contamination inside a respirator after its removal • Acute exposure to 40 DAC-h after incorporating any respiratory protection factor • Any unplanned suspected intake DOE-STD-1121-2008 59 5.6 BIOASSAY FOR DECLARED PREGNANT WORKER DOE has published a chapter in its Radiation Protection Programs Guide for Use with 10 CFR 835, Evaluation and Control of Radiation Dose to the Embryo/Fetus (DOE 2011a). All relevant parts of this document shall be used in design and operation of the parts of a bioassay program that apply to declared pregnant workers. This Technical Standard does not summarize the recommendations of that implementation guide but does note a few points about internal dosimetry. The dose limit for a declared pregnant worker’s embryo-fetus is substantially more restrictive than those for radiological workers, except for the fact that the 500-mrem limit applies to the equivalent dose for the nine-month gestation period, and not the committed equivalent dose for 50 years following intake. The maternal intakes that would cause a 500-mrem gestation period dose to the embryo-fetus are in the nominal microcurie range (e.g., for an acute inhalation at time of conception, the intakes would be approximately 1μCi for 238Pu, Type M, 50 μCi for 137Cs, Type F, and 50 μCi for 90Sr, Type M (ICRP 2002b). Routine bioassay programs designed to monitor workers should be easily adequate to demonstrate compliance with the embryo-fetus dose limits. As a verification, it may be desirable to obtain a special bioassay upon receipt of a pregnancy declaration, with a follow-up special bioassay at the conclusion of pregnancy if the worker continues to be exposed to possible intakes. ICRP Publication 88 (ICRP 2002b) provides methods for embryo-fetus internal dosimetry.

5.7 CONFIRMATORY BIOASSAY PROGRAM A confirmatory bioassay program involves limited surveillance of workers to provide verification that routine bioassay is not required. Guidance on selecting a representative sample of exposed personnel for confirmatory bioassay programs is given in NCRP Report No. 87. This sampling is based on representative air samples and the monthly time-weighted average (TWA) concentration of the radionuclide. If the TWA is, for example, 10% or less of the derived air concentration (DAC), and if the largest result used in the calculation of the TWA is less than 30% of the DAC, only a representative number of exposed personnel should participate in the bioassay program. This sample should include the most highly exposed or potentially exposed within a given area and should include at least 10% of the workers (for a total number of workers of 100 or more). If the time-weighted monthly average is very much greater than 10% of the DAC, and/ or the largest result used in the calculation is greater than 30% of the DAC, all personnel working in the area should be monitored. Individuals identified as participating in a confirmatory bioassay program in work control documents (such as RWPs) shall participate with similar rigor as those participating in bioassay programs required per 10 CFR 835.402. Failures to participate in such programs could be considered failures to implement written procedures per 10 CFR 835.104. 5.8 TIMELY RECEIPT OF BIOASSAY RESULTS Bioassay measurement results should be provided in a manner timely to the purpose for which they are obtained. Factors to consider in determining timeliness include: • use of results to implement or determine efficacy of dose reduction therapy • use of results for preliminary assessments for rapid reporting to the worker and management and for determining appropriate follow-up activities • need to confirm a suspected intake based on a high routine measurement before detection capability is lost due to normal biokinetics DOE-STD-1121-2008 60 • trade-offs in sensitivity (due to analytical short-cuts and reduced counting times) for rapid results. Because in vivo measurement data is usually available almost immediately upon completion of the measurement, the response times discussed in this section will generally apply to excreta bioassay measurements. Confirmatory bioassay measurements are not expected to show any significant detection of nuclides of concern. Since the purpose of these measurements is merely to provide general information that significant intakes are not occurring and that radiological controls are effective, the time between obtaining a bioassay sample (or measurement) and receipt of the results need not be rapid. Likewise, where routine periodic measurements are not likely to show significant intakes with regard to dose control and work administration, a 1- or 2-month analytical response time is not likely to have any significant impact. Generally speaking, a 1-month turnaround time for routine excreta sample analysis does not pose serious problems for either analytical laboratories or worker monitoring programs. Special bioassay measurements should have much faster response time. This is particularly important if the results are being used to determine need for, or efficacy of, dose reduction therapy. Rapid availability of special results is also needed for preliminary intake and dose assessments used to classify intakes for reporting purposes. It is suggested that some kind of preliminary bioassay measurements should be available within 24 to 48 hours following intake. The need for precision and accuracy in these early assessments is much less than for the measurements which will be used for the final dose assessment.

Provisions for assuring that a worker has received the appropriate in vivo measurements or has provided the scheduled excreta sample should not be overlooked in designing a program. A reasonable grace period is appropriate to deal with workers who forget to submit excreta samples or who are unable to meet the schedule. For some routine sampling frequencies, a grace period of 30 or 60 days may be appropriate. However, administrative actions (e.g., work restriction) may be appropriate for a worker who is substantially overdue for measurement. Ideally, results of new-hire or baseline measurements should be available before a worker commences the work requiring the bioassay. This prevents loss of baseline information if a sample is lost during analysis. However, loss during analysis tends to be a rare occurrence, and it is an acceptable practice to begin work once the sample has been collected but prior to receipt of results. Where air sample results form the basis for identifying intakes and making preliminary dose assessments, some kind of initial results (e.g., gross alpha or gross beta concentration) should be available within a few hours of obtaining the sample. This is particularly important for samples used to monitor for unknown or changing work conditions. Routine air samples for well-established processes and facilities may have longer turnaround times (e.g., as much as a few days), provided they are not the sole method of detecting off-normal workplace conditions. DOE-STD-1121-2008 61 6 DETECTION AND CONFIRMATION OF INTAKES Two fundamentally different kinds of signals may indicate the possibility that intakes of radionuclides have occurred. Most often, possible intakes may be indicated by workplace monitoring results (CAM alarms, survey and frisking results, air sample results) or observations (an accident, explosion, spill, leak, equipment failure). Possible intakes are more rarely signaled first by unexpected, elevated bioassay results. In some events, there is no question that intakes were possible, so special bioassay procedures and investigations are initiated to confirm or rule out intakes. In situations where the possibility is less clear, the suspicion of an intake should be investigated, that is, efforts should be made to confirm or rule out intakes, if preliminary results indicate the possibility for a significant dose. If preliminary results do not indicate the possibility of a dose above the IL of 100 mrem, then a dose may simply be assigned without investigation. A suspected intake based solely on workplace monitoring data cannot be confirmed in the same sense that repeated bioassay measurements can confirm an intake of radionuclides. There are, however, some checks that can be used to help validate the result. This is particularly important for larger predicted intakes. For example, one can look at coworker BZ data, evidence of concomitant external contamination, job-specific air monitoring information, and results of nasal smears. None of these sources provides confirmation, but collectively they can sometimes help flesh out the details of the exposure. 6.1 USE OF WORKPLACE MONITORING DATA FOR DETECTING AND CONFIRMING INTAKES

The identity of radionuclides inadvertently taken into the body and the amount of intake may be inferred using workplace monitoring data (e.g., airborne radioactivity concentration measurements, nasal- smear activity measurements, application of resuspension factors to measured surface contamination levels, etc.). Airborne radioactive material concentration data may be used as a direct indication of intake, especially if information on particle size distribution can be obtained. Evaluation of other workplace indicators proved to be useful in identifying possible intakes. However, there is no generally accepted quantitative method for correlating such indicators with intake. Heid and Jech (1972) concluded from review of several plutonium inhalation cases that the amount of activity on a nasal smear collected shortly after intake was about the same as the amount deposited in the deep lung for nose breathers and about half the deposited activity for mouth breathers. However care must be taken when evaluating potential doses based on nasal smear results where relatively small numbers of high-specific-activity plutonium particles are airborne, such as during some accidents, such as the glove box failure at Los Alamos in March 2000 that led to eight workers very briefly inhaling airborne plutonium. Intake (based on nasal swipes) was highly variable (one worker had a nasal smear which was almost 100,000 dpm, while 4 workers had smears with no detectable readings). Large variability in radioactivity intake is more the rule than the exception. This also applies to scenarios when plutonium particles penetrate respirators (Arden 2003). Brodsky (1980) suggested that a resuspension factor could be applied to surface contamination levels to assess the corresponding airborne radioactivity levels. Detailed methods are not described due to the provisional acceptance of dose assessments based on workplace monitoring data. The facility shall establish a protocol as part of the internal dosimetry program and document it in the technical basis documentation if use of workplace monitoring data appears to be appropriate for dose assessment. DOE-STD-1121-2008 62 6.2 USE OF BIOASSAY DATA FOR DETECTING AND CONFIRMING INTAKES According to the Internal Dosimetry Program chapter of the Radiation Protection Programs Guide for Use with 10 CFR 835 (DOE 2011a), intakes of radioactive materials that are suspected on the basis of a single bioassay measurement must be confirmed by one of several means. “False alarms” based on erroneous bioassay results carry a heavy penalty in terms of cost, paperwork, and public relations for both DOE and its contractors. The decision to confirm an intake based on bioassay measurements currently uses a statistical comparison of one or more results with an appropriate blank. Guidance from a variety of sources (including ANSI/HPS N13.30-2011) uses the concept of an appropriate blank for comparison with analytical measurements such as those that form the basis for bioassay measurements. However, two distinct decisions are confounded by the current method: the first is the decision whether radioactivity is above background levels, and the second is a decision whether radioactivity is above what would be expected from nonoccupational exposures, as explained in the IDG. For example, it is well known that environmental exposures to natural uranium occur, and that these have been mistaken for occupational exposures.

Figure 3. Reference Levels for Interpreting or Responding to Intake Monitoring Results Is bioassay or air monitoring datum... Record result> LDV? Is intake confirmed? > DIL? Calculate E50 Use default assumptions LDV Derived Verification Level DIL Derived Investigation Level LDMR Derived Medical Referral Level No Yes No Yes > LDMR? No Yes No Yes Record and Report Investigate Calculate HT,50 Involve Medical Staff DOE-STD-1121-2008 63 6.2.1 Decisions Based on Individual Monitoring Data Examples of actions taken following acquisition of a result from an individual monitoring program are shown in Figure 3. This approach is useful to consider, with individual sites determining the values of their respective reference levels. 6.3 STATISTICAL METHODS FOR CONFIRMING THAT AN INTAKE HAS OCCURRED Beyond the methods described in the IDG, at least two other statistical methods exist for confirming that an intake has occurred. The first is to simply pool the n bioassay results statistically to achieve the 1/ n improvement in the decision level (Strom and McGuire, 1992; Hickey et al. 1993). The second is to employ Bayesian statistical inference (Miller et al. 1993, 1995, 2000; Miller et al. 2000, 2001, 2002a, 2002b; Inkret and Miller 1995; http://www.lanl.gov/bayesian/), based on Bayes’ theorem (Bayes, 1763; Lindley 1972, 1980, 1985; Martz and Waller 1982; Calvin 1989; Press 1989). The Bayesian formalism is attractive because it incorporates prior knowledge in addition to the results of a given measurement, and it results in a distribution of likely outcomes rather than merely a point estimate with an uncertainty. However, the method has been criticized as being too subjective. At present, the DOE and this Technical Standard have taken no official position on the use of the Bayesian method. The appropriateness of Bayesian methods must be decided on a case-by-case basis. http://www.lanl.gov/bayesian/ DOE-STD-1121-2008 64 7 INTERNAL DOSE EVALUATION Radiation protection guides are expressed in terms of limiting values of dose to workers. As summed with whole body equivalent dose, 10 CFR 835.202 limits committed effective dose for individuals and committed equivalent dose for their organs and tissues. Committed effective dose and committed organ equivalent dose are calculated for intakes in specific calendar years to evaluate conformance with limiting values for occupational exposure and for reporting doses to workers. A committed effective dose is calculated 1) to evaluate conformance with limiting values for control of the workplace, 2) to measure the effectiveness of the facility’s radiation protection programs, and 3) to provide a summary to the worker of the equivalent dose that may be received in subsequent years as a result of any intake during the calendar year. The need may also arise to calculate doses over other time periods such as from the date of intake to the first year following the intake, to the date when the person would turn age 75 (i.e., “the expected lifetime dose”), and to the date of death. There are three conceptually distinct methods to assess internal dose: • assessment of intake directly from air samples or other workplace data, followed by the assessment of dose from intake • assessment of intake from bioassay data and biokinetic models, followed by the assessment of dose from intake • direct assessment of dose time-integrated retention from bioassay data, with assignment of a

putative intake that is consistent with the dose. Assessments of internal dose using mathematical biokinetic models shall be based, as appropriate, on • direct, in vivo measurements of a radionuclide(s) in various source organs of the body; or • indirect, in vitro measurements on excreta. If bioassay data are not available or are of questionable value, assessments of inhaled radionuclides should be based on workplace data, preferably on air sample measurements. The initial assessment of a radionuclide intake or retained quantity may be based on air monitoring or other workplace measurement data as well as available bioassay measurement data. However, assessments based only on workplace monitoring data should be regarded as provisional and should be updated if and when bioassay measurement data of sufficient quality become available. Evaluations of equivalent dose resulting from an intake of a radionuclide proceed from an assessment of the amount of the radionuclide in organs and tissues of the body as a function of time. The radionuclide distribution and retention depends on the physical and chemical forms of the radionuclide, its radiological properties, the physiological characteristics of the individual, the route(s) of intake, and the magnitude of intake(s). Except for radon, thoron, and their short-lived progeny, internal equivalent dose is defined in terms of the energy imparted to target tissues from the radiations emitted by radionuclides in source organs and tissues of the body. The purpose for analyzing radionuclide intake and retained quantity as a function of time is to identify the organs and tissues in which the radionuclide is deposited and to evaluate the cumulated activity (e.g., transitions in Bq∙s or μCi-days; 1 μCi-day = 3.1968E9 transitions) in source organs. Since it is often difficult to precisely determine the cumulated activity in all source organs DOE-STD-1121-2008 65 directly from bioassay measurements, biokinetic models have been developed to describe empirical relationships between intake, number of transitions, and bioassay measurement values. Bioassay and other supporting data can often require considerable expense and effort to obtain. It is neither necessary nor cost-effective to assess all intakes using the same level of effort; rather, it is more reasonable to employ a graded approach to bioassay collection and dose assessment whereby the level of effort expended on the assessment increases with the magnitude of the anticipated dose. Minor exposures may be assessed using generalized biokinetic models for a reference individual and conservative (or default) assumptions regarding the nature of the exposure and characteristics of the contaminant. The generalized model and assumptions should be based on previous experience or supporting studies at the facility or models recognized by ICRP or NCRP. The facility shall document the default models and assumptions and when these are appropriate for use. For projected doses of increased magnitude, sufficient bioassay and source characterization data shall be obtained to enable adjustments to be made to the generalized models, as appropriate, to account for the specific behavior of the radionuclide(s) in the body. The facility shall establish and document specific dose levels which require enhancement of data collection and individual specific dose assessment efforts. 7.1 DOSES TO BE ASSESSED 10 CFR 835 requires that the following doses be calculated:

• committed effective dose from intakes occurring during the year • committed equivalent dose to tissues of concern from intakes occurring during the year • total effective dose • cumulative total effective dose. The RadCon Standard also recommends the calculation of “lifetime occupational dose,” which is taken to be the same as cumulative total effective dose. 7.1.1 Committed Effective Dose In the 2007 amendment to 10 CFR 835 a provision was added to 10 CFR 835.702(b) regarding the recording of internal dose (committed effective dose or committed equivalent dose). The provision allows for not recording of any positive monitoring result estimated to correspond to an individual receiving less than 0.01 rem (0.1 mSv) committed effective dose. Typically, this would be for very sensitive bioassay protocols or for radionuclides which are easy to detect at very low doses (e.g., routine tritium bioassay). The bioassay or air monitoring result used to make the estimate shall be maintained in accordance with 10 CFR 835.703(b) and the unrecorded internal dose estimated for any individual in a year shall not exceed the applicable monitoring threshold at 10 CFR 835.402(c). All confirmed occupational intakes, above the decision level (i.e., the statistical decision level), shall be assessed. Based on each assessment, doses shall be calculated for each intake during the calendar year for positive monitoring results corresponding to doses greater than 0.01 rem (0.1 mSv) committed effective dose E50. Where there are multiple intakes or where several radionuclides are involved, each facility may establish a per-radionuclide or a per-intake minimum assessment value so that the intent of the above DOE-STD-1121-2008 66 recommendation is met. 7.1.2 Committed Equivalent Dose to Tissue of Concern Each facility shall identify the tissues of concern relative to radionuclides at the facility and shall justify and document the selection in the technical basis document. Specific requirements for identifying the organs of concern are given in the definition of weighting factor in 10 CFR 835. Wound site tissue and associated lymph nodes should be excluded from committed equivalent dose calculations (Nénot and Stather 1979; National Research Council 1988). The committed equivalent dose HT,50 to the tissue(s) of concern shall be calculated for those years where a committed effective dose is calculated. For exposures to the short-lived progeny of radon and thoron, Hlung,50 may be calculated as E50 divided by the tissue weighting factor for lung, wlung = 0.12. 7.1.3 Total Effective Dose Total effective dose shall be calculated in cooperation with the site’s external dosimetry program and records program pursuant to 10 CFR 835. Total effective dose includes all occupational doses: internal, external, and those received at other sites. An operational form of total effective dose is given by the IAEA (1996) in its Basic Safety Standards as T p , ing , ing , inh , inh( ) (50) (50) ,j j j j j j Total Effective Dose E H d e I e I= = + +∑ ∑ where ET is the total effective dose, Hp(d) is the deep dose equivalent (i.e., the equivalent dose to the whole body is assess at a depth of 1 cm in tissue), the summations are over j radionuclides, the e(50)s are committed effective dose per unit intake coefficients for ingestion and inhalation from ICRP-68 (1994b) for the appropriate physical and chemical forms of radionuclides j, and the Is are the intakes of the radionuclides via ingestion or inhalation. The terms of the summation can be calculated using codes such as IMBA Expert™ USDOE-Edition or equivalent codes.

7.1.4 Cumulative Total Effective Dose Committed doses from intakes prior to January 1, 1989, may be included in lifetime dose calculations. Including such doses gives a more accurate estimate of the lifetime accumulation and is consistent with the recommendations of NCRP Reports 91 and 116 (1987 and 1993). However, to demonstrate compliance with 10 CFR 835 requirements, these doses should be kept separate from the cumulative total effective dose from intakes occurring after January 1, 1989. 7.2 DATA NEEDS AND DEFAULT ASSUMPTIONS Generally, the more data available, the more precise the dose determination. However, practical considerations generally limit the amount of data available. Internal dosimetry programs should commit resources in proportion to the magnitude of potential doses. For doses below the IL, it is acceptable to use default assumptions as described in the technical basis documentation. DOE-STD-1121-2008 67 7.3 INTERPRETATION OF BIOASSAY DATA Selection of methods for bioassay interpretation plays an important role in the design of the bioassay program. For example, in cases where either the intake scenario or the biological retention cannot be well known, more bioassay data are needed to adequately arrive at the dose estimation. Conversely, if the intake, uptake, and retention models are well characterized and apply to the exposure scenario, one bioassay measurement which confirms a previous result may be sufficient for dose assessment. Since there is normally sufficient uncertainty in both the bioassay data and the biokinetic models, the use of multiple data points and fitting to the model may be necessary. Facility-specific and radionuclide-specific decisions about bioassay interpretation methods shall be documented and shall dictate a significant part of the overall bioassay and internal dosimetry program. The derivation of intakes and retained quantities from bioassay data may be the critical step in the dose assessment process. Evaluations of exposure to internal radionuclides should account for all possible sites of retention and their associated retention times (if known) in the body. Generalized biokinetic models, suitably modified to account for experience or studies at the facility, may provide a starting point for the initial assessment of an intake and for determining the specific needs for follow-up bioassay measurements. All organs contributing to the effective dose, calculated with the weighting factors given in 10 CFR 835, shall be considered rather than only those organs in which the radionuclide can be readily measured. 7.3.1 Direct Estimation of Retained Quantity When thorough bioassay histories are attainable, and good confidence can be placed on organ and whole body radionuclide content evaluations, it is possible to explicitly derive the retained quantity and retention history of an exposure without resorting to use of default parameters. In some cases the uncertainties associated with the biokinetics are much greater than the uncertainty in the direct assessments of intake and retained quantity. A tritium exposure with sufficient urine assay data to document the biological excretion rate is an example of using excretion history. Due to uncertainty in the route of intake (e.g., skin absorption versus inhalation) and in the biological clearance rate (which depends on water consumption), the tritium excretion history provides the best assessment of the number of transitions and, thus, the equivalent dose. Similarly, a radioiodine exposure, well documented in time and monitored by in vivo thyroid counting, can be assessed directly from the bioassay result. In both cases, discrete or parameterized methods of summation of transitions in the well-known source organs will provide sufficient information for dose assessment. Where direct uptake and retention history are used for dose assessment, the method for converting data to equivalent dose shall be documented as part of the dose assessment. However, if the bioassay data are insufficient for a thorough assessment of retained quantities, or are of such poor quality that whole body or pertinent organ content cannot be directly derived, then biokinetic models shall be used. 7.3.2 Biokinetic Modeling A biok