ML14336A637

DG-8031 December, 2014 Draft Regulatory Guide 8.34 Monitoring Criteria and Methods to Calcluate Occupational Radiation Doses

ML14336A637
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Issue date:	12/08/2014
From:	Steven Garry NRC/NRR/DRA
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Garry S
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RG-8.034 DG-8031, RG-8.034, Rev. 1
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U.S. NUCLEAR REGULATORY COMMISSION December 2014 OFFICE OF NUCLEAR REGULATORY RESEARCH Revision 1 DRAFT REGULATORY Technical Lead Steve Garry GUIDE 1 DG-8031 December, 2014 2 DRAFT REGULATORY GUIDE 8.34 3 (First Draft was issued as DG-8031, on October 2013) 4 5 MONITORING CRITERIA AND METHODS TO 6 CALCULATE OCCUPATIONAL RADIATION DOSES 7 A. INTRODUCTION 8

9 Purpose 10 This guide provides methods acceptable to the staff of the U.S. Nuclear Regulatory 11 Commission (NRC) for monitoring the occupational radiation dose to individuals and for 12 calculating occupational radiation doses. The regulatory guide (RG) applies to both reactor and 13 materials licensees under both NRC and Agreement State licenses.

14 Applicable Rules and Regulations 15 The regulations established by the NRC in Title 10, Energy, of the Code of Federal 16 Regulations (10 CFR) Part 20, Standards for Protection against Radiation (Ref. 1),

17 Section 20.1101, Radiation Protection Programs, establish requirements for licensees (a) to keep 18 individuals exposures to radiation below the specified regulatory radiation dose limits and (b) to 19 keep such radiation doses as low as is reasonably achievable (ALARA). To demonstrate 20 compliance with the dose limits, licensees must perform surveys and, when appropriate, monitor 21 individuals radiation exposure and calculate the doses resulting from the exposure.

22 Also, 10 CFR 20.1201, Occupational Dose Limits for Adults, establishes radiation dose 23 limits for occupationally exposed individuals. These limits apply to the sum of the dose received 24 from external exposure and the dose from internally deposited radioactive material. Conditions 25 that require individual monitoring of external and internal occupational doses are specified in 26 10 CFR 20.1502, Conditions Requiring Individual Monitoring of External and Internal 27 Occupational Dose. Monitoring the intake of radioactive material and assessing the committed 28 effective dose equivalent (CEDE) (for internal exposures) is required by 10 CFR 20.1502(b). The 29 calculations that licensees are required to perform in order to comply with these regulations were 30 affected by the 2007 revisions of 10 CFR 20.1003 and 10 CFR 50.2 (Ref. 2), both titled Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Website under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.

Electronic copies of this regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRCs public Web site under Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collection/. The regulatory guide is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under ADAMS Accession No. MLXXXXXXX. The regulatory analysis may be found in ADAMS under Accession No. MLXXXXXXXXX and the staff responses to the public comments on DG-8031 may be found under ADAMS Accession No. MLXXXXXXX.

31 Definitions. This revision redefined the total effective dose equivalent (TEDE) as the sum of 32 the effective dose equivalent (for external exposures) and the CEDE (for internal exposures).

33 The following regulatory requirements are also discussed in this guide:

34

• 10 CFR 20.1007, Communications 35 36

• 10 CFR Part 19, Notices, Instructions, and Reports to Workers: Inspection and 37 Investigations(Ref. 3) 38 39

• 10 CFR 20.1202, Compliance with Requirements for Summation of External and 40 Internal Doses 41 42

• 10 CFR 20.1204, Determination of Internal Exposure 43 44

• 10 CFR 20.1206, Planned Special Exposures 45 46

• 10 CFR 20.1207, Occupational Dose Limits for Minors 47 48

• 10 CFR 20.1208, Dose Equivalent to an Embryo/Fetus 49 50

• 10 CFR 20.1501, General, in 10 CFR 20 Subpart F, Surveys and Monitoring 51 52

• 10 CFR 20.1703, Use of Individual Respiratory Protection Equipment 53 54

• 10 CFR 20.2106, Records of Individual Monitoring Results 55 56

• 10 CFR 20.2206, Reports of Individual Monitoring 57 Related Guidance 58 The NRC has developed guidance related to calculating occupational doses for monitored 59 individuals and has provided criteria regarding which individuals should be monitored for radiation 60 exposure. Such guidance includes the following:

61

• RG 8.7, Instructions for Recording and Reporting Occupational Radiation 62 Exposure Data (Ref. 4) 63 64

• RG 8.9, Revision 1, Acceptable Concepts, Models, Equations, and Assumptions 65 for a Bioassay Program (Ref. 5) 66 67

• RG 8.11, Applications of Bioassay for Uranium (Ref. 6) 68 69

• RG 8.25, Revision 1, Air Sampling in the Workplace (Ref. 7) 70 71

• RG 8.29, Instruction Concerning Risks from Occupational Radiation Exposure 72 (Ref. 8) 73 RG 8.34, Revision 1, Page 2

74

• RG 8.35, Revision 1, "Planned Special Exposures (Ref. 9) 75 76

• RG 8.36, Radiation Dose to the Embryo/Fetus (Ref. 10) 77 78

• RG 8.40, Methods for Measuring Effective Dose Equivalent from External 79 Exposure (Ref. 11) 80 81 82 Purpose of Regulatory Guides 83 The NRC issues RGs to describe to the public methods that the staff considers acceptable 84 for use in implementing specific parts of the agencys regulations, to explain techniques that the 85 staff uses in evaluating specific problems or postulated accidents, and to provide guidance to 86 applicants. RGs are not substitutes for regulations and compliance with them is not required.

87 Methods and solutions that differ from those set forth in RGs will be deemed acceptable if they 88 provide a basis for the findings required for the issuance or continuance of a permit or license by the 89 Commission.

90 Paperwork Reduction Act 91 This RG discusses information-collection requirements covered by 10 CFR Part 20 and 92 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, that the Office of 93 Management and Budget (OMB) approved under OMB control numbers 3150-0014 94 and 3150-0011 respectively. The NRC may neither conduct nor sponsor, and a person is not 95 required to respond to, an information-collection request or requirement unless the requesting 96 document displays a currently valid OMB control number.

97 RG 8.34, Revision 1, Page 3

98 TABLE OF CONTENTS 99 A. INTRODUCTION ..................................................................................................................... 1 100 PURPOSE ....................................................................................................................................... 1 101 APPLICABLE RULES AND REGULATIONS...................................................................................... 1 102 RELATED GUIDANCE .................................................................................................................... 2 103 PURPOSE OF REGULATORY GUIDES ............................................................................................. 3 104 PAPERWORK REDUCTION ACT ..................................................................................................... 3 105 B. DISCUSSION ............................................................................................................................ 5 106 REASON FOR REVISION ................................................................................................................ 5 107 BACKGROUND .............................................................................................................................. 5 108 OCCUPATIONAL DOSE LIMITS FOR ADULTS, MINORS, AND EMBRYOS/FETUSES ........................ 6 109 PLANNED SPECIAL EXPOSURES (PSES) ....................................................................................... 6 110 SURVEYS ...................................................................................................................................... 6 111 MONITORING AT LEVELS SUFFICIENT TO DEMONSTRATE COMPLIANCE .................................... 7 112 USE OF EFFECTIVE DACS ............................................................................................................ 7 113 ALPHA MONITORING AT NUCLEAR POWER PLANTS ................................................................... 7 114 DISCRETE RADIOACTIVE-PARTICLE MONITORING AND SDE ...................................................... 8 115 HARMONIZATION WITH INTERNATIONAL STANDARDS................................................................ 9 116 DOCUMENTS DISCUSSED IN STAFF REGULATORY GUIDANCE .................................................... 9 117 C. STAFF REGULATORY GUIDANCE.................................................................................. 10 118 1. MONITORING CRITERIA ..................................................................................................... 10 119 2. OCCUPATIONAL DOSE ........................................................................................................ 11 120 3. PROSPECTIVE ASSESSMENTS OF THE NEED FOR OCCUPATIONAL DOSE MONITORING ..... 11 121 4. DETERMINATION OF EXTERNAL DOSES ............................................................................. 12 122 5. DETERMINATION OF INTAKES ............................................................................................ 14 123 6. DETERMINATION OF INTERNAL DOSES .............................................................................. 15 124 7. USE OF INDIVIDUAL OR MATERIAL-SPECIFIC INFORMATION ............................................ 18 125 8. LIMITATION ON URANIUM INTAKE .................................................................................... 18 126 9. RECORDING OF INDIVIDUAL MONITORING RESULTS ........................................................ 18 127 D. IMPLEMENTATION ............................................................................................................ 19 128 REFERENCES............................................................................................................................. 20 129 APPENDIX A, METHODS OF CALCULATING INTERNAL DOSE ................................. 23 130 131 132 133 134 135 RG 8.34, Revision 1, Page 4

136 B. DISCUSSION 137 138 Reason for Revision 139 This revision of RG 8.34 provides updated regulatory guidance on monitoring criteria and 140 methods of calculating occupational dose based on the revised definition of the TEDE. This RG 141 also provides updated guidance on acceptable methods of:

142

• Determining the need for monitoring and demonstrating compliance with occupational 143 dose limits.

144

• Monitoring alpha intakes and determining internal dose from alpha-emitting radionuclides.

145

• Assessing deep-dose equivalent (DDE) when the measurements of the primary monitoring 146 device (dosimeter) are inconsistent with other radiological measurements (e.g., surveys or 147 electronic dosimeters).

148

• Assessing intakes and committed dose equivalent (CDE) from wounds.

149

• Examples of calculational methods to assess intakes and internal doses.

150 Background 151 On December 4, 2007, the NRC revised the definition of the TEDE in 10 CFR 20.1003 and 152 10 CFR 50.2 (as published in the Federal Register at 72 FR 68043 (Ref. 12)). The revision 153 subsequently affected the methods of monitoring and calculating occupational radiation doses and 154 demonstrating compliance with the occupational dose limits. Previously, the definition of the 155 TEDE was the sum of the DDE (to account for external exposure) and the CEDE (to account for 156 internal exposure). Under the revised rule 10 CFR 20.1003, the TEDE was redefined by replacing 157 the DDE with the effective dose equivalent-external (EDEX).

158 Old definition: TEDE = DDE + CEDE 159 New definition: TEDE = EDEX + CEDE 160 Regulations in 10 CFR 20.1201(c) require that, when external exposure is determined by 161 measurement with an external personal monitoring device, the DDE for the part of the body 162 receiving the highest exposure be used in place of the effective dose equivalent (i.e., the EDEX) 163 unless the EDEX is determined by a dosimetry method approved by the NRC (see RG 8.40). In 164 uniform radiation fields, the EDEX is normally determined by measuring the DDE and, therefore, 165 the revised TEDE definition has little impact on monitoring methods. However, for exposures in 166 non-uniform radiation fields, the revised TEDE definition provides greater monitoring flexibility 167 and accuracy for licensees in monitoring worker exposures. Under non-uniform conditions, the 168 previous TEDE definition tended to provide dose assessments that were excessively conservative.

169 Occupational dose limits are applicable during routine operations, planned special exposures, 170 and during emergencies. Doses received during declared nuclear emergencies (including 171 international emergencies) must be included in the determination of annual occupational dose.

172 However, the potential for exceeding a dose limit during a declared emergency should not prevent a 173 licensee from taking necessary actions to protect health and safety.

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174 175 Occupational Dose Limits for Adults, Minors, and Embryos/Fetuses 176 For adults, occupational dose limits (except for planned special exposures) are established in 177 10 CFR 20.1201(a) as follows:

178

• For protection against stochastic effects, the annual TEDE limit is 5 rem 179 (50 millisieverts (mSv)).

180

• For protection against nonstochastic effects, the annual total organ dose equivalent 181 (TODE) limit is 50 rem (500 mSv).

182

• For protection of the lens of the eye, the annual lens dose equivalent (LDE) limit is 183 15 rem (150 mSv).

184

• For protection of the skin of the whole body or of the skin of any extremity, the 185 annual shallow-dose equivalent (SDE) limit is 50 rem (500 mSv).

186 For minors, occupational dose limits are established in 10 CFR 20.1207, Occupational Dose 187 Limits for Minors, as annual limit at 10 percent of the adult dose limits.

188 For the embryo/fetus of a declared pregnant woman, a dose equivalent limit during the entire 189 pregnancy is established in 10 CFR 20.1208, Dose Equivalent to an Embryo/Fetus, as 0.5 rem 190 (5 mSv).

191 Planned Special Exposures (PSEs) 192 PSEs are subject to the conditions specified in 10 CFR 20.1206, Planned Special 193 Exposures (e.g., exceptional circumstances, specific authorizations, and informing and instructing 194 the worker). RG 8.35, Planned Special Exposures, provides guidance on conducting PSEs. For 195 dose-accounting purposes, dose received during a PSE is in addition to and accounted for 196 separately from the dose that is limited by 10 CFR 20.1201.

197 Surveys1 198 Surveys (i.e., evaluations of the radiological conditions and potential hazards) should be 199 conducted as necessary in support of radiological monitoring and calculation of occupational dose.

200 Instruments and equipment used in performing surveys must be calibrated periodically for the type 201 of radiation measured in accordance with 10 CFR 20.1501(c).

202 When a licensee assigns or permits the use of respiratory protection equipment to limit the 203 intake of radioactive material, 10 CFR 20.1703(c)(2) requires surveys and bioassays, as necessary, 204 to evaluate actual intakes. Indications of an intake could include facial contamination, nasal 1 Survey means an evaluation of the radiological conditions and potential hazards incident to the production, use, transfer, release, disposal, or presence of radioactive material or other sources of radiation. When appropriate, such an evaluation includes a physical survey of the location of radioactive material and measurements or calculations of levels of radiation or concentrations or quantities of radioactive material present.

RG 8.34, Revision 1, Page 6

205 contamination, malfunctioning respiratory protection equipment, loss of engineering controls 206 creating an airborne radioactivity area, and work in unknown or unplanned airborne radioactivity 207 areas.

208 During operations, licensees should perform airborne radioactivity surveys as required in 209 10 CFR 20.1502 to characterize the radiological hazards that may be present and, as appropriate, 210 use engineering and respiratory protection equipment to reduce intakes. When it is not practical to 211 use process or engineering controls to reduce the concentrations of airborne radioactivity to values 212 below those that define an airborne radioactivity area, licensees are required under 213 10 CFR 20.1702(a), to be consistent with keeping the TEDE ALARA, to increase monitoring 214 (e.g., perform air sampling and track Derived Air Concentration (DAC)-hours and bioassay 215 measurements) and to limit intakes by using access controls, limiting exposure times, or having 216 individuals use respiratory protection equipment.

217 Monitoring at Levels Sufficient To Demonstrate Compliance 218 Regulations in 10 CFR 20.1502 require monitoring at levels sufficient to demonstrate 219 compliance with the occupational dose limits; therefore, monitoring methods should be reasonably 220 accurate. In addition, licensees may voluntarily issue individual monitoring devices or use 221 calculational methodologies for reasons other than for required personnel monitoring under the 222 requirements in 10 CFR 20.1502 (e.g., to inform individuals of exposure conditions, or to alleviate 223 safety concerns). The results of monitoring that is voluntarily provided but not required by 224 10 CFR 20.1502 are not subject to the dose recording or reporting requirements in 10 CFR Part 20, 225 Subpart L, Records, or Subpart M, Reporting. However, licensees may voluntarily provide these 226 reports to the exposed individual(s) and to the NRC.

227 Use of Effective DACs 228 The regulation at 10 CFR 20.1204(e) provides a method for determining internal exposure 229 when the identity and concentration of each radionuclide in a mixture is known. The identities and 230 concentrations of radionuclides may be determined based on representative radiological surveys 231 identifying the specific radionuclides and quantifying their relative mix. Once the relative mix is 232 known, licensees may apply scaling factors applicable to the mixture for use in calculating DACs and 233 tracking DAC-hours as specified in 20.1204(e). This is commonly referred to as effective DACs 234 and is applicable to beta/gamma activity, alpha activity, and hard-to-detect radionuclides.

235 The use of effective DAC values may be needed in operational radiological protection 236 programs to establish airborne radioactivity postings, determining alarm set points for continuous air 237 monitors, determining the need for respiratory protection, estimating internal dose, or determining 238 when bioassay measurements may be needed. When using effective DACs, licensees may disregard 239 those radionuclides in the mixture (based on prior representative surveys) having a concentration less 240 than 10% of the radionuclides DAC, given that the sum of disregarded radionuclides does not exceed 241 30% (see 10 CFR 20.1204(g)).

242 243 Alpha Monitoring at Nuclear Power Plants 244 For reactor facilities that have experienced significant fuel defects, alpha contamination 245 may be a radiological hazard requiring specific evaluation. Alpha contamination (when present)

RG 8.34, Revision 1, Page 7

246 requires specific evaluation because the DAC values for alpha emitting isotopes are generally 247 orders of magnitude more restrictive than DACs for beta-emitting and gamma-emitting isotopes.

248 Each facility should characterize and update its alpha source term as needed based on the 249 facilitys operational history. Alpha source-term characterization should not be based solely on the 250 samples of dry activated waste collected for waste-classification purposes under 10 CFR Part 61, 251 Licensing Requirements for Land Disposal of Radioactive Waste. Loose contamination surveys 252 may not be sufficient to identify fixed alpha contamination that may pose a hazard during abrasive 253 work (e.g., grinding, cutting, or welding). The characterization should determine the extent of the 254 alpha hazard within the facility such as within localized areas.

255 The extent of the radiological characterization that is needed depends on the relative 256 significance of the alpha source term compared to other radiological contaminants. The 257 characterization may be used to determine the specific alpha radionuclides and to determine their 258 relative concentrations in a mixture. Once the relative concentrations are known, an effective DAC 259 may be determined and used in radiological protection and dose assessment (in lieu of using the 260 most restrictive DAC of any radionuclide in the mixture as required by 10 CFR 20.1204(f)).

261 The principal transuranic nuclides producing alpha radiological hazards include the 262 isotopes of curium, plutonium, and americium. For historical fuel failures (e.g., ten years have 263 passed since significant fuel failure), the shorter-lived curium-242 will have largely decayed, 264 leaving the longer-lived alpha radionuclides with more restrictive DACs and annual limits on 265 intake (ALI) as the most prevalent hazard. However, investigations of more recent fuel failures are 266 likely to identify curium-242 as the most abundant alpha-emitting nuclide, which has less 267 restrictive DAC and ALI values. Therefore, effective DAC values must be updated as needed to 268 account for the time-dependent (decayed) mix of alpha radionuclides. In addition, consideration 269 should be given to transuranic isotopes which decay by other than alpha emission (e.g., Pu-241).

270 The extent of radiological protection measures against alpha radionuclides may be 271 determined based on:

272

• knowledge of the specific alpha radionuclide mix 273

• knowledge of the solubility/insolubility of the radionuclides 274

• conservative assumptions about the most restrictive radionuclide in the mixture 275

• determination of site-specific effective-DAC alpha values 276 Discrete Radioactive-Particle Monitoring and SDE 277 A discrete radioactive particle (DRP) is a small (usually microscopic) and highly 278 radioactive particle emitting either only beta or both beta and gamma radiation and having 279 relatively high specific activity. DRPs are primarily an external exposure hazard to the skin, as 280 measured by the SDE.

281 In 2002, the NRC amended its regulations related to the shallow-dose equivalent/skin-dose 282 limit in 10 CFR Part 20 (at 67 FR 16298 (Ref. 13); see also Regulatory Issue Summary 2002-10, 283 Revision of the Skin Dose Limit in 10 CFR Part 20 (Ref. 14)). The amended regulations 284 changed the definition and method of calculating SDEs by specifying that the assigned SDE must 285 be the dose averaged over the contiguous 10 cm2 of skin receiving the highest exposure.

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286 Harmonization with International Standards 287 The NRC has a goal of harmonizing its guidance (to the extent that this is practical) with 288 international standards. The International Commission on Radiological Protection (ICRP) and the 289 International Atomic Energy Agency (IAEA) have issued a significant number of standards, 290 guidance and technical documents, and recommendations addressing good practices in most 291 aspects of radiation protection. The NRC encourages licensees to consult the international 292 documents noted throughout this guide and implement the applicable good practices they contain 293 that are consistent with NRC regulations.

294 Such documents include the following:

295

• ICRP Publication 26, Recommendations of the International Commission on 296 Radiological Protection (Ref. 15) 297

• ICRP Publication 30, (7-volume set including supplements), Limits for Intakes of 298 Radionuclides by Workers (Ref. 16) 299

• ICRP Publication 54, Individual Monitoring for Intakes of Radionuclides by Workers 300 (Ref. 17) 301

• ICRP Publication 60, 1990 Recommendations of the International Commission on 302 Radiological Protection (Ref. 18) 303

• ICRP Publication 68, Dose Coefficients for Intakes of Radionuclides for Workers 304 (Ref. 19) 305

• ICRP Publication 78, Individual Monitoring for Internal Exposure of Workers (Ref. 20) 306

• ICRP Publication 103, The 2007 Recommendations of the International Commission on 307 Radiological Protection (Ref. 21) 308 Documents Discussed in Staff Regulatory Guidance 309 Although this RG uses information, in part, from one or more reports developed by 310 external organizations and other third-party guidance documents, the RG does not endorse these 311 references other than as specified in this RG. These reports and third-party guidance documents 312 may contain references to other reports or third-party guidance documents (secondary 313 references). If a secondary reference has itself been incorporated by reference in NRC regulations 314 as a requirement, licensees and applicants must comply with that requirement in the regulation.

315 If the secondary reference has been endorsed in an RG as an acceptable approach for 316 meeting an NRC requirement, the reference constitutes a method acceptable to the NRC staff for 317 meeting that regulatory requirement as described in the specific RG. If the secondary reference has 318 neither been incorporated by reference in NRC regulations nor endorsed in an RG, the secondary 319 reference is neither a legally binding requirement nor a generic NRC approval as an acceptable 320 approach for meeting an NRC requirement. However, licensees and applicants may consider and 321 use the information in the secondary reference, if it is appropriately justified and consistent with 322 current regulatory practice, in ways consistent with applicable NRC requirements such as those in 323 10 CFR Part 20.

324 RG 8.34, Revision 1, Page 9

325 C. STAFF REGULATORY GUIDANCE 326 327 1. Monitoring Criteria 328 329 Regulations in 10 CFR 20.1502 require individual monitoring of external and internal 330 occupational dose at levels sufficient2 to demonstrate compliance with the occupational dose 331 limits. As a minimum, licensees must monitor occupational exposure to radiation from licensed and 332 unlicensed radiation sources3 under the control of the licensee.

333 334 For external occupational exposure, licensees are required to supply and require the use of 335 individual monitoring devices if the external occupational dose:

336 337

• for adults, is likely to exceed 10 percent of the occupational dose limits in 338 10 CFR 20.1201(a);

339

• for minors, in one year, is likely to exceed a deep-dose equivalent of 0.1 rem (1 mSv),

340 a lens dose equivalent of 0.15 rem (1.5 mSv), or a shallow-dose equivalent to the skin 341 of the whole body or to the skin of the extremities of 0.5 rem (5 mSv); or 342

• for declared pregnant women, during their entire pregnancy, is likely to exceed a 343 deep-dose equivalent of 0.1 rem (1 mSv), and 344

• for individuals entering a high or very high radiation area.

345 346 For internal occupational exposure, licensees are required to monitor the intake of 347 radioactive material and assess the CEDE by 10 CFR 20.1502(b) if the intake is likely to exceed:

348

• 10 percent of the applicable annual limit on intake (ALI) for adults; 349

• 0.1 rem (1 mSv) for minors in one year; or 350

• 0.1 rem (1 mSv) for declared pregnant women during the entire pregnancy.

2 Monitoring performed to assess the magnitude of an inadvertent or unplanned exposure (from external radiation or from intakes of radionuclides) is required monitoring per 10 CFR 20.1502 (i.e., required to demonstrate compliance with the dose limits in Part 20) and are subject to the recording requirements in 20.2106(a) and the reporting requirements 20.2206(b).

3 Unlicensed sources are radiation sources not licensed by the NRC or Agreement States; such as products or sources covered by exemptions from licensing requirements (e.g., 10 CFR 30.14, Exempt Concentrations; 10 CFR 30.15, Certain Items Containing Byproduct Material; 10 CFR 30.18, Exempt Quantities; 10 CFR 30.19, Self-Luminous Products Containing Tritium, Krypton-85, or Promethium-147; 10 CFR 30.20, Gas and Aerosol Detectors Containing Byproduct Material; 10 CFR 30.22, Certain Industrial Devices; or 10 CFR 40.13, Unimportant Quantities of Source Material), naturally occurring radioactive materials that are not covered by the Atomic Energy Act, radioactive materials possessed by or nuclear facilities operated by another Federal entity such as the U.S. Department of Defense or the U.S. Department of Energy, and machines that produce radiation (such as x-ray radiography machines and x-ray machines used by security staff).

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351 2. Occupational Dose 352 The definition of occupational dose in 10 CFR 20.1003 includes dose received during the 353 course of employment in which assigned duties involve exposure to radiation or radioactive 354 material from licensed and unlicensed sources of radiation, whether in the possession of the 355 licensee or of another person. The definition of occupational dose was changed in 1995 (at 356 60 FR 36038) (Ref. 22) so that occupational dose applies to workers whose assigned duties involve 357 exposure to radiation, irrespective of their location inside or outside a restricted area. Note:

358 A member of the public does not become an occupationally exposed individual simply as a result of 359 entering a restricted area.

360 Individuals who receive occupational exposure and are likely to receive more than 361 100 mrem must be instructed in accordance with 10 CFR 19.12, Instruction to Workers. See 362 RG 8.29 for further information.

363 3. Prospective Assessments of the Need for Occupational Dose Monitoring 364 Licensees must identify those individuals receiving occupational dose, either individually 365 or as a group or category of individuals. Individuals pre-designated by the licensee as receiving 366 occupational dose are subject to the occupational dose limits; otherwise, individuals must be 367 considered as members of the public subject to public dose limits in 10 CFR 20.1301, Dose Limits 368 for Individual Members of the Public.

369 Once occupationally exposed individuals are identified, licensees should perform a 370 prospective assessment to determine whether those individuals are likely to exceed the minimum 371 exposure levels specified in 10 CFR 20.1502 (i.e., to determine the need for monitoring of the 372 occupational dose). The potential for unlikely exposures and accident conditions need not be 373 considered because these events, by definition, are unlikely. However, as discussed at 374 60 FR 36039, the term likely to receive includes normal situations as well as abnormal 375 situations involving exposure to radiation which can reasonably be expected to occur during the life 376 of the facility. Therefore, licensees should consider normal operations and anticipated operational 377 occurrences (e.g., unplanned onsite events, such as sudden increases in external radiation levels, or 378 localized areas of high airborne radioactivity) but would not need to consider design-basis 379 accidents 380 The prospective assessment determines the type of monitoring required (e.g., external-dose or 381 internal-dose monitoring). In performing a prospective assessment, an evaluation should be 382 performed based on planned work activities and likely exposure conditions. In the prospective 383 assessment, licensees may take credit for the use of engineering controls (e.g., containment, 384 decontamination, ventilation, and filtration). However, if licensees are using respiratory protection 385 equipment to limit the intake of radioactive material, licensees must establishing a respiratory 386 protection program and perform air sampling, surveys, and bioassays to evaluate intakes and 387 estimate dose in accordance with the 10 CFR 20.1703. Prospective assessments should be revised 388 when there are substantial changes to the radiological conditions of personnel exposure 389 (e.g., changes in work activities, airborne concentrations, beta energy spectra, or use of 390 radiation-producing equipment emitting new or different types of energies).

391 392 The requirements for monitoring in 10 CFR 20.1502 refer to exposures that might occur at 393 each licensee individually. Doses that have already been received while in the employ of another 394 licensee, or that might be received in the future while in the employ of another licensee or 395 unlicensed entity, are excluded from consideration in a licensees determination of the need to RG 8.34, Revision 1, Page 11

396 monitor an individual. The need for monitoring should be based on the anticipated exposure to 397 licensed or unlicensed sources under the control of a single licensee.

398 4. Determination of External Doses 399 a. Determination of the TEDE 400 Under 10 CFR 20.1202, if a licensee is required to monitor both external dose and internal 401 dose, the licensee must demonstrate compliance with the dose limits by summing external and 402 internal doses (i.e., TEDE = EDEX + CEDE). However, if the licensee is required to monitor only 403 external doses under 10 CFR 20.1502(a) or only internal doses under 10 CFR 20.1502(b),

404 summation is not required to demonstrate compliance with the occupational dose limits. For 405 example, if the internal dose is not monitored, the CEDE can be assumed to be equal to zero and the 406 TEDE is equal to the EDEX. Similarly, if the external dose is not monitored, the EDEX can be 407 assumed to be equal to zero and the TEDE is equal to the CEDE.

408 b. Determination of the EDEX 409 The EDEX is determined using one or more combinations of the following methods in 410 accordance with 10 CFR 20.1201(c). These methods are described in RG 8.40 as follows:

411 1. Measuring the DDE at the most highly exposed part of the whole body with an external 412 personal monitoring device, as required by 10 CFR 20.1201(c), when an NRC method for 413 determining EDEX is not used.

414 2. Measuring external exposure with one or more external personal monitoring devices and 415 determining EDEX using an NRC-approved method (such as those provided in RG 8.40 or 416 as specifically approved elsewhere by the NRC).

417 3. Calculating the EDEX based on survey data obtained under 10 CFR 20.1501 or on other 418 radiological data (such as known source activity, dose rates, and exposure times) using 419 scientifically sound technical methods. This might be required (a) under unique exposure 420 situations (e.g., if an individuals body were partially exposed to radiation streaming in a 421 narrow beam geometry), (b) when the individuals monitoring device was not in the region 422 of the highest whole-body exposure (in accordance with 10 CFR 20.1201(c)), or (c) when 423 the results of the individual monitoring are not available (i.e., the monitoring device is 424 damaged or lost).

425 Note: Within the same monitoring period, a licensee may use a combination of the 426 methods above: A licensee may routinely determine EDEX for the majority of a monitoring period 427 using method 1 above, and then use method 2 or 3 for special exposure situations at other times.

428 The results of the different dosimetry methods must be combined to determine the EDEX for the 429 entire monitoring period.

430 c. Determination of the Deep-Dose Equivalent (DDE) 431 The DDE (external exposure of the whole body) is typically measured with a passive 432 primary monitoring device that assesses the dose at a tissue depth of 1 centimeter (cm) (a mass 433 thickness of 1,000 mg/cm2). The DDE can also be calculated if the appropriate parameters are 434 known (i.e., the radiation source strength, the exposure geometry, and whether full or partial 435 shielding was in place).

RG 8.34, Revision 1, Page 12

436 An individual monitoring device located at the most highly exposed part of the whole body 437 measuring the DDE is a conservative and (for uniform exposures) a reasonably accurate estimate of 438 the EDEX. However, if the radiation dose is highly non uniform, causing a specific part of the 439 whole body (head, trunk, arms above the elbow, or legs above the knees) to receive a substantially 440 higher dose than the rest of the whole body, the individual monitoring device should be placed near 441 that part of the whole body expected to receive the highest dose. There are several other 442 NRC-approved methods for determining EDEX provided in RG 8.40.

443 444 In many exposure situations, a required monitoring device (e.g., a passive dosimeter) may 445 be voluntarily supplemented with an additional, active dosimeter (e.g., an electronic dosimeter 446 used for work control and daily dose accounting purposes). Due to the differences in dosimeter 447 design and detection technology, and the relative measurement errors associated with each type of 448 dosimeter, there can be valid differences in readings of these two dosimeters for the same exposure, 449 even if the dosimeters are co-located on the monitored individual. Within a reasonable, licensee 450 pre-determined accuracy criteria (depending on dosimeter designs), small differences between 451 measurements can be disregarded and either dosimetry value used as the measured dose (since both 452 results are considered valid and equal within measurement error). However, a significantly higher 453 reading on the voluntary dosimeter may indicate that the required dosimeter was not appropriately 454 placed to measure the highest exposed part of the whole body. Licensees should investigate those 455 cases where a significant discrepancy exists between dosimeters. If the differences cannot be 456 resolved, an assessment must be performed to determine the DDE, LDE, and SDE for the highest 457 exposed part of the whole body, as provided for in 10 CFR 20.1201(c).

458 d. Determining the LDE 459 If the LDE is being monitored with a dosimeter, that dosimeter should be calibrated to 460 measure the dose at a tissue depth of 0.3 centimeter (cm) (a mass thickness 300 mg/cm2).

461 Alternatively, the LDE may be conservatively determined based on SDE measurements at 462 7 mg/cm2. In many exposure situations, safety glasses can be worn to minimize exposures to the 463 lens of the eye from low-energy (or poorly penetrating) radiations, potentially eliminating the need 464 for monitoring the LDE.

465 e. Determination of the SDE 466 The SDE is defined only for external exposure at a tissue depth of 0.007 cm (a mass 467 thickness of 7 mg/cm2), and is the dose averaged over the contiguous 10 cm2 of skin receiving the 468 highest exposure. If the SDE is being measured with a dosimeter, that dosimeter should be 469 calibrated to measure the dose at a tissue depth of 7 mg/cm2. For skin contamination, the computer 470 code described in NUREG/CR-6918, VARSKIN: A Computer Code for Skin Contamination 471 Dosimetry (Ref. 23) may be used to assess the SDE. The SDE may also be determined from 472 analytical calculational methods based on survey data when dosimetry methods are not 473 representative of the actual exposure conditions.

474 The SDE for exposure to submersion-class radionuclides containing low-energy betas is 475 not readily measurable by direct survey techniques or dosimetry methods and hence may need to be 476 calculated based on air-sample analyses and DAC-hr tracking. This submersion exposure 477 information may be needed for informing workers of radiological exposure conditions 478 (e.g., informing workers of the SDE rates during pre-job briefings) and also to account in dose 479 records for the SDE that might not be adequately measured by dosimeters (e.g., because of the 480 dosimeters lack of response to a low-energy beta spectrum).

RG 8.34, Revision 1, Page 13

481 5. Determination of Intakes 482 For those licensees monitoring internal dose in accordance with 10 CFR 20.1204, a 483 determination must be made of the intake that can occur through inhalation, ingestion, absorption 484 through the skin, or absorption through wounds. The amount of the intake may be assessed from 485 suitable and timely measurements of airborne radionuclides or may be based on bioassay 486 measurements.

487 The assessment of intake should include not only the readily detected radionuclides but 488 also the hard-to-detect radionuclides if their dose contribution is significant. The activity of 489 hard-to-detect radionuclides may be based on scaling factors that correspond to the amount of 490 readily detected radionuclides. See RG 8.25, Air Sampling in the Workplace, and Regulatory 491 Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program, 492 for further guidance on determining uptakes and intakes.

493 Unless respiratory protection is used, the concentration of radionuclides in the intake 494 (i.e., the breathing-zone concentration) is assumed to be equal to the ambient concentration.

495 Therefore, when selecting the air-sample location, one should consider engineered features such as 496 containment, airflow, and filtration to ensure that the air sample is representative of the air 497 breathed.

498 If respiratory protection is used to limit the intake of radioactive materials, 499 10 CFR 20.1703(c)(4)(i) requires internal monitoring to be implemented as part of the respiratory 500 protection program. When respiratory protection is provided, the intake is adjusted by dividing the 501 ambient air concentration by the appropriate Assigned Protection Factor (APF) listed in 502 Appendix A, Assigned Protection Factors for Respirators, to 10 CFR 20. If the ambient air 503 concentration is determined by performing breathing-zone air sampling inside the respiratory 504 protective device (such as with a lapel air sampler inside a loose-fitting supplied air hood or suit),

505 no APF adjustment is made to the ambient air concentration as measured in the breathing-zone air 506 sample.

507 a. Determining the Intake Based on Air Sampling 508 Intake (I) based on air-sampling results can be assessed by multiplying the airborne 509 concentration (C) by the breathing rate and the exposure time:

510 I = CAir sample (µCi/ml)

• breathing rate (ml/minutes)

• exposure time (minutes), where the 511 breathing rate of a Reference Man under light working conditions is 2E+4 ml/minute 512 (20 liters/minute).

513 The intake of radionuclides can also be estimated by DAC-hour tracking in which the 514 ambient airborne concentration (expressed as a fraction of the DAC) is multiplied by exposure time 515 (expressed in hours).

516 If the intake assessment is based on measurements from a lapel air sampler, the intake may 517 be assessed by multiplying the activity on the lapel air sampler by the breathing rate divided by the 518 lapel air samplers flow rate as follows:

519 I = AAir sample (µCi)

• breathing rate/air sampler flow rate (ml/min), where the breathing rate 520 of a Reference Man under light working conditions is 2E+4 ml/minute 521 (20 liters/minute).

RG 8.34, Revision 1, Page 14

522 b. Determining the Intake Based on Bioassay Measurements 523 The intake can be determined based on initial bioassay measurements of uptakes and on 524 follow-up bioassay measurements to determine the retention/elimination rates (which can also 525 assist in the evaluation of the mode of intake (inhalation or ingestion)). Time and motion 526 conditions may support assessments of intake as well. Guidance on methods of estimating intake 527 based on bioassay measurements of uptake is provided in NUREG/CR-4884, Interpretation of 528 Bioassay Measurements (Ref. 24).

529 Any intake from wounds is generally assessed based on bioassay measurements using a 530 combination of whole body in vivo bioassay and handheld instrumentation. The bioassay 531 measurements should determine the location and depth of the injected source so that CDE dose 532 calculations may be made to the most highly exposed 10 cm2 area of the skin at a depth of 0.007 cm 533 (see Section 6.d below).

534 Note: The amount of the intake may be assessed using newer, updated biokinetic models 535 (e.g., those described in ICRP Publication 60, 1990 Recommendations of the International 536 Commission on Radiological Protection, and ICRP Publication 103, The 537 2007 Recommendations of the International Commission on Radiological Protection). However, 538 the CEDE must be calculated using the existing 10 CFR 20.1003 organ weighting factors (unless 539 the use of other weighting factors has been specifically approved by the NRC).

540 c. Determining Intakes of Alpha Emitters 541 Alpha intakes may be assessed based on gross surface area and/or airborne surveys of the 542 alpha-emitting isotopes present in the work area at the time of exposure. Scaling factors based on 543 beta/gamma activity may be determined and used to assess the identity and relative concentration 544 of alpha isotopes.

545 Internal doses may also be assessed based on whole-body count data and scaling factors 546 when nominal (e.g., less than 500 mrem CEDE) alpha doses occur. However, when an alpha intake 547 resulting in alpha doses exceeding a nominal quantity is considered likely, excreta sampling or lung 548 counting may be needed to assess intakes and assign dose. When excreta sampling is to be 549 initiated, sampling should begin as soon as possible following detection of the exposure and should 550 continue for a 24-hour period or until at least one sample is collected (following the first void for 551 urine). ANSI N13.39-2001 (R2011), Design of Internal Dosimetry Programs (Ref. 25), provides 552 additional guidance on excreta sampling.

553 6. Determination of Internal Doses 554 a. Calculation of the Committed Effective Dose Equivalent (CEDE) 555 The dose quantity for protection against stochastic effects of internal dose is the CEDE; 556 i.e., a 50-year committed effective dose equivalent from intakes occurring during the monitoring 557 period. There are three fundamental methods described below for calculating the CEDE:

RG 8.34, Revision 1, Page 15

558

• Using dose coefficients4 from the U.S. Environmental Protection Agencys Federal 559 Guidance Report No. 11 (FGR-11) (Ref. 26).

560

• Using ALI methods.

561

• Using DAC-hour methods.

562 For details about and examples of calculating the CEDE, see Appendix A.

563 Note: When performing CEDE calculations using the ALI and DAC-hour methods, the 564 ALI and DAC values provided in Appendix B, Annual Limits on Intake (ALIs) and Derived Air 565 Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; 566 Concentrations for Release to Sewerage, to 10 CFR Part 20 must be used unless the licensee has 567 obtained prior NRC approval in accordance with 10 CFR 20.1204(c)(2) to adjust the ALI or DAC 568 values.

569 b. Calculation of the Committed Dose Equivalent (CDE) 570 The CDE is the 50-year committed dose equivalent from the intake of radioactive material.

571 For methods and examples of calculating the CDE, see Appendix A. The special case of 572 calculating the CDE from wound intakes is discussed in Section 6.d below.

573 c. Calculation of the Total Organ Dose Equivalent (TODE) 574 The dose limit for protection against nonstochastic effects is expressed in terms of the 575 TODE; i.e., the sum of the DDE and the CDE.

576 TODE = DDE + CDE 577 The TODE is determined by adding the DDE (measured at the most highly exposed part of 578 the whole body) to the CDE.

579 If only internal monitoring is being performed, the TODE is equal to the CDE to the most 580 highly exposed organ (given that the DDE was not monitored and is assumed to be equal to zero).

581 Further details on acceptable methods of calculating the CDE are described in Appendix A.

582 If both internal and external monitoring are being performed, the licensee must 583 demonstrate that both the 5-rem TEDE and the 50-rem TODE limits are met. One method of 584 demonstrating compliance with the TODE limit is by summing the DDE and the CDE to the most 585 highly exposed organ. Another acceptable method of demonstrating that the TODE limit is met is 586 by keeping the maximum DDE below 5 rem and the CEDE below 1 rem5; if this is done, the TODE 587 cannot exceed its 50-rem limit. In this case, the CDE does not need to be determined because 4

Note: Federal Guidance Report No. 11 (FGR-11) uses the terminology dose conversion factors. However, more recent ICRP documents use the terminology dose coefficients. This regulatory guide is adopting the newer terminology dose coefficients (this change in terminology is acceptable because the terminology is not incorporated in the regulations).

5 The value of 1 rem is based on the most limiting tissue-weighting factor (i.e., the weighting factor for the thyroid tissue is 0.03; therefore, 1 rem divided by thyroid weighting factor of 0.03 results in a CDE of 33.3 rem. A CDE value of 33.3 rem, when added to an assumed 5-rem DDE value, is less than the CDE limit of 50 rem.

RG 8.34, Revision 1, Page 16

588 compliance was demonstrated by calculation. If the CEDE does exceed 1 rem, the CDE must be 589 determined in order to demonstrate compliance with the dose limits.

590 d. Doses from Intakes through Wounds 591 In accordance with 10 CFR 20.1202(d), the licensee shall evaluate and, to the extent 592 practical, account for intakes through wounds.

593 Regulations in10 CFR 20.1201 also specify two annual dose limits:

594

• TODE limits (10 CFR 20.1201(a)(1)(ii))the sum of the DDE and the CDE to any 595 individual organ or tissue other than the lens of the eye being equal to 50 rem 596 (0.5 Sv)and 597

• SDE limits (10 CFR 20.1201(a)(2)(ii))the SDE to the skin of the whole body or skin of 598 any extremity being equal to 50 rem (0.5 Sv).

599 However, because the SDE is defined only for external exposure, the SDE quantity and its 600 dose limit are not applicable to dose from wound intakes. Therefore, the TODE dose limit becomes 601 the only applicable limit; i.e., a CDE limit of 50 rem to any individual organ, including the skin.

602 Note that in most skin-exposure situations, the skin dose is from external exposure (and therefore 603 the dose to the skin is normally equal to the SDE). However, when the dose to the skin is from a 604 wound, the CDE dose limit applies (not the SDE).

605 In making the TODE dose calculation (to the skin organ) under 20.1201(a)(1)(ii), the DDE 606 component is zero (because DDE is specifically defined as an external whole-body exposure). As a 607 result, the CDE is determined for the basal layer of the skin at a depth of 0.007 cm below skin 608 surface for the most highly exposed, contiguous 10-cm2 area.

609 In summary, the CDE to the skin is the appropriate quantity to be calculated as the 610 integrated dose from the time of injection to the time the source is removed or by the 50-year 611 integration period for committed dose. The CDE is to be determined at a depth of 612 0.007 centimeters below the surface of the skin, averaged over the most highly exposed 10 cm2 of 613 the basal layer of the skin. In order to do this calculation, the location (depth) of the source and 614 distance to the basal layer must be determined as an input parameter. The VARSKIN computer 615 code may be used in performing the CDE skin-dose calculations.

616 Bioassay measurements should be performed to determine whether there is a systemic 617 uptake from the injected radioactive material. For wound intakes with systemic uptakes, an 618 evaluation must be performed of the CEDE and TEDE. Additional information on assessing 619 intakes through wounds is available in ICRP-54, ICRP-78, NCRP-87 (Ref. 27), and technical 620 articles by Toohey (Ref. 28) and Ishigure (Ref. 29).

621 Note: With respect to tissue dose, there is no regulatory limit for small-volume localized 622 tissue dose. However, licensees should estimate the committed dose to small volumes of 623 underlying tissues (e.g., 1 cm3) at the wound site for purposes of determining the potential for 624 tissue impairment and whether medical intervention is warranted (e.g., surgical removal). The 625 guidance in National Council on Radiation Protection & Measurements (NCRP) Report No. 156, 626 Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and Procedures for 627 Their Assessment, Dosimetry, and Treatment (Ref. 30), is acceptable for this evaluation.

628 e. Calculating the CDE and CEDE for Inhalation, Submersion and Absorption RG 8.34, Revision 1, Page 17

629 A number of methods are acceptable for calculating the CDE and CEDE from the intake of 630 radioactive materials. Some of these methods are described below. However, calculations of the 631 CEDE must be based on organ weighting factors and tissues specified in 10 CFR Part 20. The dose 632 coefficients based on ICRP Publication 60 cannot be used unless specifically approved by the 633 NRC, because ICRP 60 and ICRP 103 tissues and weighting factors are different from those in 634 10 CFR Part 20.

635 7. Use of Individual or Material-Specific Information 636 The regulation in 10 CFR 20.1204(c) states that when specific information on the 637 physical and biochemical properties of the radionuclides taken into the body or the behavior of the 638 material in an individual is known, the licensee may [...] use that information to calculate the 639 committed effective dose equivalent [...]. Prior NRC approval is not required, but detailed records 640 must be kept to demonstrate the acceptability of the dose assessment.

641 The characteristics most amenable to such individual or site-specific consideration are the 642 activity median aerodynamic diameter (AMAD) of the inhaled aerosol and the solubility (or 643 insolubility) of the material in the lungs and in the gastrointestinal (GI) tract (particularly for alpha 644 intakes). The use of specific information on the physical and biochemical properties to calculate 645 the CEDE requires the licensee to do considerably more work and to have greater technical 646 expertise than the other methods, so this method might not be useful for small infrequent intakes.

647 Conversely, the use of specific information on the physical and biochemical properties of 648 radionuclides taken into the body might be appropriate in the cases of accidental large exposures if 649 more accurate information would lead to a better estimate of the actual dose.

650 8. Limitation on Uranium Intake 651 In accordance with 10 CFR 20.1201(e), in addition to the annual dose limits, the licensee 652 shall limit the soluble uranium intake by an individual to 10 mg in a week, in consideration of its 653 chemical toxicity. RG 8.11, Applications of Bioassay for Uranium, describes methods 654 acceptable for the design of bioassay programs for protection against intake of uranium, conditions 655 under which bioassay is necessary, minimum quantifiable values for direct and indirect bioassay 656 measurements, protection guidelines, and objectives.

657 9. Recording Of Individual Monitoring Results 658 The requirements for recording individual monitoring results are contained in 10 CFR 20.2106, 659 which requires that the recording be done on NRC Form 5, or in clear and legible records 660 containing all the information required by NRC Form 5. Regulatory Guide 8.7 provides further 661 guidance for recording and reporting occupational radiation dose data.

662 663 Licensees should avoid entering doses on NRC Form 5 with more significant figures than justified 664 by the precision of the basic measured values. In general, it is appropriate to enter dose values with 665 two significant figures on NRC Form 5 using the standard rules for round-off. Thus, a 666 computer-generated calculated dose of "1.726931 rems" should be entered on NRC Form 5 as "1.7 667 rems." However, licensees should generally carry at least three significant figures in calculations to 668 avoid loss of accuracy due to multiple round-offs.

669 670 In addition, licensees should not enter doses smaller than 0.001 rem on NRC Form 5 because 671 smaller values are insignificant relative to the dose limits. Therefore, a calculated committed RG 8.34, Revision 1, Page 18

672 effective dose equivalent of "0.006192 rem" should be entered as "0.006 rem," and a value of 673 "0.000291 rem" should be entered as "0 rem."

674 D. IMPLEMENTATION 675 676 The purpose of this section is to provide information to applicants and licensees regarding 677 the NRCs plans for using this RG.

678 Methods or solutions that differ from those described in this regulatory guide may be 679 deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that 680 the proposed alternative complies with the appropriate NRC regulations. Current licensees may 681 continue to use guidance the NRC found acceptable for complying with the identified regulations 682 as long as their current licensing basis remains unchanged.

683 RG 8.34, Revision 1, Page 19

684 REFERENCES6 685 686 1. U.S. Code of Federal Regulations (CFR), Standards for Protection against Radiation, 687 Part 20, Chapter I, Title 10, Energy.

688 2. 10 CFR 50, Domestic Licensing of Production and Utilization Facilities, Part 50, 689 Chapter I, Title 10, Energy.

690 3. 10 CFR 19, Notices, Instructions, and Reports to Workers: Inspection and 691 Investigations, Part 19, Chapter I, Title 10, Energy.

692 4. U.S. Nuclear Regulatory Commission (NRC), Instructions for Recording and Reporting 693 Occupational Radiation Exposure Data, RG 8.7, Revision 2, November 2005, 694 Agencywide Documents Access and Management System (ADAMS) Accession 695 No. ML052970092.

696 5. NRC, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay 697 Program, RG 8.9, Revision 1, July 1993, ADAMS Accession No. ML003739554.

698 6. NRC, Applications of Bioassay for Uranium, RG 8.11, June 1974, ADAMS Accession 699 No. ML003739450.

700 7. NRC, Air Sampling in the Workplace, RG 8.25, Revision 1, June 1992, ADAMS 701 Accession No. ML003736916.

702 8. NRC, Instruction Concerning Risks from Occupational Radiation Exposure, RG 8.29, 703 Revision 1, February 1996, ADAMS Accession No. ML003739438.

704 9. NRC, "Planned Special Exposures, RG 8.35, Revision 1, August 2010, ADAMS 705 Accession No. ML101370008.

706 10. NRC, Radiation Dose to the Embryo/Fetus, RG 8.36, July 1992, ADAMS Accession 707 No. ML003739548.

708 11. NRC, Methods for Measuring Effective Dose Equivalent from External Exposure, 709 RG 8.40, July 2010, ADAMS Accession No. ML100610534.

710 12. NRC, Occupational Dose Records, Labeling Containers, and the Total Effective Dose 711 Equivalent, Federal Register, Vol. 72, No. 232, December 4, 2007, pp. 68043-68059 712 (72 FR 68043).7 6

Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html The documents can also be viewed online or printed for a fee in the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or (800) 397-4209; fax (301) 415-3548; or e-mail pdr.resource@nrc.gov.

7 Printed copies of Federal Register notices are available for a fee from the U.S. Government Printing Office, 732 N. Capitol Street NW, Washington, DC 20401, telephone (866) 521-1800, or they may be downloaded for free from the Government Printing Office Web site, http://www.gpo.gov/fdsys/.

RG 8.34, Revision 1, Page 20

713 13. NRC, Revision of the Skin Dose Limit, Federal Register, Vol. 67, No. 66, April 5, 2002, 714 pp. 16298-16301 (67 FR 16298).

715 14. NRC, Revision of the Skin Dose Limit in 10 CFR Part 20, Regulatory Issue 716 Summary 2002-2010, July 9 2002, ADAMS Accession No. ML021860332.

717 15. International Commission on Radiological Protection (ICRP), Recommendations of the 718 International Commission on Radiological Protection, ICRP Publication 26, Oxford, UK:

719 Pergamon Press, 1977.

720 16. ICRP, Limits for Intakes of Radionuclides by Workers, ICRP Publication 30 (7-volume 721 set including supplements), Oxford, UK: Pergamon Press, 1982.

722 17. ICRP, Individual Monitoring for Intakes of Radionuclides by Workers, ICRP 723 Publication 54, Oxford, UK: Pergamon Press, 1989, specifically Sections 4.2 and 4.3.

724 18. ICRP, 1990 Recommendations of the International Commission on Radiological 725 Protection, ICRP Publication 60, Oxford, UK: Pergamon Press, 1990.

726 19. ICRP, Dose Coefficients for Intakes of Radionuclides for Workers, ICRP 727 Publication 68, Oxford, UK: Pergamon Press, 1994.

728 20. ICRP, Individual Monitoring for Internal Exposure of Workers, ICRP Publication 78, 729 Oxford, UK: Pergamon Press, 1997, specifically Section 4.2.

730 21. ICRP, The 2007 Recommendations of the International Commission on Radiological 731 Protection, ICRP Publication 103, Oxford, UK: Pergamon Press, 2007.

732 22. NRC, Radiation Protection Requirements: Amended Definitions and Criteria, Federal 733 Register, Vol. 60, No. 134, July 13, 1995, pp. 36038-36043 (60 FR 36038).

734 23. NRC, VARSKIN 5: A Computer Code for Skin Contamination Dosimetry, 735 NUREG/CR-6918, Rev. 2, July 2014, Accession No. ML14204A361.

736 24. NRC, Interpretation of Bioassay Measurements, NUREG/CR-4884, June 1990, 737 ADAMS Accession No. ML11285A018.

738 25. American National Standards Institute (ANSI), Design of Internal Dosimetry Programs, 739 ANSI N13.39-2001 (R2011), Washington, DC, 2011.

740 26. Eckerman, K.F., A.B. Wolbarst, and A.C.B. Richardson, Limiting Values of 741 Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, 742 Submersion, and Ingestion, Federal Guidance Report No. 11 (EPA 520/1-8-020),

743 U.S. Environmental Protection Agency, Washington, DC, 1988.

744 27. National Council on Radiation Protection & Measurements (NCRP), Use of Bioassay 745 Procedures for Assessment of Internal Radionuclide Deposition, NCRP Report No. 87, 746 Bethesda, MD, March 1987, specifically subsections 5.3.1, 5.3.2, and 5.4.6.

747 28. Toohey, R.E., et al., Dose Coefficients for Intakes of Radionuclides via Contaminated 748 Wounds, Health Physics 100(5):508-14, May 2011; a much larger and revised version 749 (Ver. 2, August 2014) is available from the Oak Ridge Institute for Science and Education RG 8.34, Revision 1, Page 21

750 at http://orise.orau.gov/reacts/resources/retention-intake-publication.aspx (accessed 751 October 10, 2014).

752 29. Ishigure, N., Implementation of the NCRP Wound Model for Interpretation of Bioassay 753 Data for Intake of Radionuclides Through Contaminated Wounds, Journal of Radiation 754 Research 50(3):267-76, May 2009.

755 30. NCRP, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds and 756 Procedures for Their Assessment, Dosimetry, and Treatment, NCRP Report No. 156, 757 Bethesda, MD, 2007.

758 31. Oak Ridge National Laboratory, ORNL/TM-13188, Recommended ALIs and DACs for 759 10 CFR 20: A Consistent Numerical Set (1996), ADAMS Accession No.

760 ML14322A420.

761 762 RG 8.34, Revision 1, Page 22

763 Appendix A 764 Methods of Calculating Internal Dose 765 766 767 1. Calculations of the CDE and the CEDE Based on Bioassay Measurements 768 Using Federal Guidance Report No. 11 (FGR-11) 769 This method is based on using tabulated dose coefficients to calculate the dose. FGR-11 770 provides tables of dose coefficients (DCs) (FGR-11 uses the terminology dose conversion factors) 771 for intakes by inhalation and by ingestion (see excerpt below for inhalation of cobalt-60 (Co-60)).

772 FGR-11 provides two types of DCs:

773 774 (1) DCs for the CDE to an organ or tissue per unit of activity (DCorgan) (e.g., the 775 heading Lung below) and 776 777 (2) DCs for the CEDE per unit of activity (DCeffective) (as shown in the far right column 778 of the tables under the heading Effective).

779 780 If site-specific information is known about the type of compound and its clearance class, the 781 appropriate clearance class can be selected. If not, the class is normally selected based on the most 782 conservative class; in Example 1, the DC for the lung is selected from clearance Class Y, which has a 783 value of 3.45E-7). Multiplying the DCs by the intake (I) for that radionuclide yields the CDE and 784 CEDE for that radionuclide.

785 786 CDE (rem) = DCorgan (rem/µCi [rem per millicurie])

• I (µCi) 787 CEDE (rem) = DCeffective (rem/Ci)

• I (Ci) 788 789 Example 1: Calculations of the CDE and the CEDE for Co-60, based on bioassay 790 measurements using the DCs from FGR-11. Note: The DCs in FGR-11 are tabulated in Sieverts 791 per Becquerel (Sv/Bq) and may be converted to millirem per microcurie (mrem/Ci) by 792 multiplying by 3.7E+9.

793 RG 8.34, Revision 1, Page 23

794 795 An intake by inhalation was estimated by a whole body count to be 360 nanocuries (nCi) 796 (0.36 µCi) of Co-60 as a Class Y aerosol. Calculate the CDE to the lung and the CEDE.

797 From Table 2.1 of FGR-11 (see excerpt below), the DCs for the Class Y Co-60 798 radionuclide are 3.45E-7 Sv/Bq for the CDE and 5.91E-8 Sv/Bq for the CEDE.

799 800 801 DClung = (3.45E-7 Sv/Bq) * (3.7E+9) = 1277 mrem/µCi 802 DCeffective = (5.91E-8 Sv/Bq) * (3.7E+9) = 219 mrem/µCi 803 804 The doses are calculated by multiplying these DCs by the intake of 0.36 µCi:

805 806 CDElung = (1277 mrem/µCi) * (0.36 µCi) = 460 mrem 807 CEDE = (219 mrem/µCi) * (0.36 µCi) = 79 mrem 808 809 2. Calculation of the CEDE based on Bioassay Measurements using Stochastic 810 ALIs 811 The ALI values are listed in Table 1 of 10 CFR 20, Appendix B, Annual Limits on Intake 812 (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; 813 Effluent Concentrations for Release to Sewerage. Column 1 lists the values for oral ingestion and 814 Column 2 lists the values for inhalation. The stochastic ALI values can be used in the calculation 815 of the CEDE, which is based on the fraction of the allowable annual intake and the 5-rem 816 (50-millisievert (mSv)) CEDE dose limit. When the ALI is defined by the stochastic limit, this 817 value alone is given in the table.

818 819 820 RG 8.34, Revision 1, Page 24

821 Because the stochastic ALI corresponds to a 5-rem (50-mSv) CEDE dose limit, the CEDE 822 may be calculated based on the ratio of the intake to the stochastic ALI multiplied by 5 rem 823 (50 mSv):

824 825 CEDE = (I/ALI)

• 5 rem 826 827 Example 2: Calculate the CEDE based on bioassay measurements using the stochastic ALI.

828 The intake by inhalation for a worker was estimated by bioassay to be 360 nCi (0.36 µCi) 829 of Co-60 as a Class Y aerosol. Calculate the CEDE.

830 831 From Appendix B above, Table 1, Column 2, the ALI for Class Y Co-60 is:

832 833 ALI (stochastic) = 30 µCi 834 CEDE = (I/ALI)

• 5 rem 835 CEDE = (0.36 µCi/30 µCi)

• 5 rem = 0.06 rem = 60 mrem 836 837 Note: Doses calculated based on FGR-11 methods are generally more precise than doses 838 calculated based on ALI values, because ALI values are given to only one significant figure.

839 Additionally, the precision of the ALI values is limited by the calculational technique used in 840 ICRP-30 (Section 4.7) whereby target organs that are not significantly irradiated were excluded 841 (<10% rule), as well as dose from source organs contributing less than 1% were also excluded. For 842 further information, see Oak Ridge National Laboratory, ORNL/TM-13188, Recommended ALIs 843 and DACs for 10 CFR 20: A Consistent Numerical Set (Ref. 31).

844 845 For Co-60, a 60-mrem value based on an ALI calculation compares to a calculated CEDE 846 value of 79 mrem using the FGR-11 method as determined in Example 1 above. For other 847 radionuclides such as Co-58, the differences might be larger. However, either calculational method 848 and/or result is acceptable in demonstrating compliance with regulatory limits.

849 850 3. Calculation of the CDE Based on Bioassay Measurements Using 851 Nonstochastic ALI 852 The 10 CFR 20 Appendix B , Table 1, Column 2, nonstochastic ALI values can be used in 853 the calculation of the CDE, based on the fraction of the allowable annual intake and the 50-rem 854 (500-mSv) CDE dose limit. When the ALI is defined by the nonstochastic limit, this value is listed 855 first in the table with its corresponding organ (see excerpt below), and the corresponding stochastic 856 ALI are given in parentheses (e.g., 9E+1 µCi (90 µCi) for ingestion and 2E+2 µCi (200 µCi) for 857 inhalation in the excerpt below).

RG 8.34, Revision 1, Page 25

858 859 860 861 Because the nonstochastic ALI corresponds to a 50-rem (500-mSv) CDE dose limit, the 862 CDE may be calculated based on the ratio of the intake to the nonstochastic ALI multiplied by 863 50 rem (500 mSv):

864 CDE = (I/ALI)

• 50 rem 865 866 Note: For a mixture of radionuclides, the sum of the fractions technique as described in 867 10 CFR 20.1202(b) must be used.

868 Example 3: Calculate the CDE based on bioassay measurements using the nonstochastic ALIs.

869 870 The intake by inhalation for a worker was estimated by bioassay to be 131 nCi (0.131 µCi) 871 of iodine-131 (I-131) as a Class D aerosol. Calculate the CDE to the thyroid.

872 873 From Appendix B above, Table 1, Column 2, the ALI for Class D I-131 is:

874 875 ALI (nonstochastic) = 5E+1 µCi = 50 µCi 876 CDE = (0.131 µCi/50 µCi)

• 50 rem = 0.131 rem = 131 mrem 877 878 4. Calculation of the CDE Based on Air Sampling and Nonstochastic 879 DAC-Hours (DAC-hr) 880 For nonstochastic radionuclides, an exposure to an airborne concentration of 1 DAC for 881 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> results in a 50-rem CDE, or 50,000 mrem/2000 hours, or a 25-mrem CDE per DAC-hr.

882 CDE = (25 mrem per DAC-hr)

• number of DAC-hr 883 where the number of DAC-hr = (air concentration / DAC value)

• exposure time.

884 Example 4: Calculate the CDE based on air sampling and nonstochastic DAC-hr.

885 886 Calculate the CDE to the thyroid for a 30-minute exposure based on an air-sample result of 887 2.1E-7 µCi/ml from I-131.

RG 8.34, Revision 1, Page 26

888 889 The nonstochastic DAC for I-131 is listed in Appendix B (see the excerpt below) as 890 2E-8 µCi/ml.

891 892 893 CDE = 25 mrem/DAC-hr * (2.1E-7 µCi/ml / 2E-8 µCi/ml) number of DACs * (0.5 hr) =

894 131 mrem 895 5. Calculations of the CEDE Based on Air Sampling and Stochastic DAC-hr 896 For stochastic radionuclides (e.g., Co-60), an exposure to an airborne concentration of 897 1 DAC results in a 5000-mrem CEDE in 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> of exposure time (5000 mrem/2000 hours) or 898 a 2.5-mrem CEDE per stochastic DAC-hr.

899 CEDE = 2.5 mrem/DAC-hr

• number of DAC-hr 900 where the number of DAC-hr = (air concentration / DAC value)

• exposure time.

901 Example 5: Calculate the CEDE based on air sampling and stochastic DAC-hr.

902 903 Calculate the CEDE for a 30-minute exposure based on an air sample result of 904 2.1E-7 µCi/ml from Co-60.

905 906 From Appendix B below, the stochastic DAC for Co-60 in a clearance Class Y compound 907 is 1E-8 µCi/ml.

908 909 RG 8.34, Revision 1, Page 27

910 CEDE = (2.5 mrem/DAC-hr) * [(2.1E-7 µCi/ml) / (1E-8 µCi/ml)] number of DACs

• 911 (0.5 hr) = 26 mrem 912 6. Calculation of the CEDE Based on Air Sampling and Calculated Stochastic 913 DAC-hr 914 CEDE = 2.5 mrem/DAC-hr

• number of DAC-hr 915 Number of DAC-hr = air concentration / calculated DAC value

• exposure time 916 Note: Appendix B to 10 CFR Part 20 does not list the stochastic DAC values (as shown in 917 the empty circled cell below) for radionuclides with intakes that have nonstochastic limits.

918 However, the stochastic DAC values may be calculated based on the stochastic ALI values. These 919 stochastic ALI values are listed (in parentheses) below the limiting nonstochastic organ (see circled 920 value of 2E+2 µCi in the table below).

921 922 Example 6: Calculate the CEDE based on air sampling and calculated stochastic DAC-hr.

923 Calculate the CEDE for a 30-minute exposure based on an air-sample result of 924 2.1E-7 µCi/ml from I-131.

925 The stochastic DAC value is first calculated by dividing the stochastic ALI by the breathing 926 rate of 2.4E+9 ml/yr.

927 The calculated stochastic DAC for I-131 = (2E+2 µCi) / (2.4E+9 ml/yr) = 8E-8 µCi/ml 928 or µCi/cc (because 1 ml = 1 cc).

929 CEDE = (2.5 mrem/hr/DAC-hr) * [(2.1E-7 µCi/ml) / (8E-8 µCi/ml)] DACs * (0.5 hr) 930 = 3.3 mrem RG 8.34, Revision 1, Page 28