ML14336A637

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

ML14336A637
Person / Time
Issue date:	12/08/2014
From:	Steven Garry NRC/NRR/DRA
To:	NRC/NRR/DRA
Garry S
Shared Package
ML14336A640	List: ML14336A637
References
RG-8.034 DG-8031, RG-8.034, Rev. 1
Download: ML14336A637 (28)

Text

U.S. NUCLEAR REGULATORY COMMISSION December 2014OFFICE OF NUCLEAR REGULATORY RESEARCH Revision 1 DRAFT REGULATORY GUIDE Technical LeadSteve Garry Written suggestions regarding this guide or development of new guides may be submitted through the NRC's 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 NRC's 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 NRC's 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.

DG-8031 December, 2014 1 DRAFT REGULATORY GUIDE 8.34 2 (First Draft was issued as DG-8031, on October 2013) 3 4 MONITORING CRITERIA AND METHODS TO 5 CALCULATE OCCUPATIONAL RADIATION DOSES 6 A. INTRODUCTION 7 8 Purpose 9 This guide provides methods acceptable to the staff of the U.S. Nuclear Regulatory 10 Commission (NRC) for monitoring the occupational radiation dose to individuals and for 11 calculating occupational radiation doses. The regulatory guide (RG) applies to both reactor and 12 materials licensees under both NRC and Agreement State licenses.

13 Applicable Rules and Regulations 14 The regulations established by the NRC in Title 10, "Energy," of the Code of Federal 15 Regulations (10 CFR) Part 20, "Standards for Protection against Radiation" (Ref. 1), 16 Section 20.1101, "Radiation Protection Programs," establish requirements for licensees (a) to keep 17 individuals' exposures to radiation below the specified regulatory radiation dose limits and (b) to 18 keep such radiation doses "as low as is reasonably achievable" (ALARA). To demonstrate 19 compliance with the dose limits, licensees must perform surveys and, when appropriate, monitor 20 individuals' radiation exposure and calculate the doses resulting from the exposure.

21 Also, 10 CFR 20.1201, "Occupational Dose Limits for Adults," establishes radiation dose 22 limits for occupationally exposed individuals. These limits apply to the sum of the dose received 23 from external exposure and the dose from internally deposited radioactive material. Conditions 24 that require individual monitoring of external and internal occupational doses are specified in 25 10 CFR 20.1502, "Conditions Requiring Individual Monitoring of External and Internal 26 Occupational Dose." Monitoring the intake of radioactive material and assessing the committed 27 effective dose equivalent (CEDE) (for internal exposures) is required by 10 CFR 20.1502(b). The 28 calculations that licensees are required to perform in order to comply with these regulations were 29 affected by the 2007 revisions of 10 CFR 20.1003 and 10 CFR 50.2 (Ref. 2), both titled 30 RG 8.34, Revision 1, Page 2 "Definitions." This revision redefined the "total effective dose equivalent" (TEDE) as the sum of 31 the effective dose equivalent (for external exposures) and the CEDE (for internal exposures).

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

33

• 10 CFR 20.1007, "Communications" 34 35

• 10 CFR Part 19, "Notices, Instructions, and Reports to Workers: Inspection and 36 Investigations"(Ref. 3) 37 38

• 10 CFR 20.1202, "Compliance with Requirements for Summation of External and 39 Internal Doses" 40 41

• 10 CFR 20.1204, "Determination of Internal Exposure" 42 43

• 10 CFR 20.1206, "Planned Special Exposures" 44 45

• 10 CFR 20.1207, "Occupational Dose Limits for Minors" 46 47

• 10 CFR 20.1208, "Dose Equivalent to an Embryo/Fetus" 48 49

• 10 CFR 20.1501, "General," in 10 CFR 20 Subpart F, "Surveys and Monitoring" 50 51

• 10 CFR 20.1703, "Use of Individual Respiratory Protection Equipment" 52 53

• 10 CFR 20.2106, "Records of Individual Monitoring Results" 54 55

• 10 CFR 20.2206, "Reports of Individual Monitoring" 56 Related Guidance 57 The NRC has developed guidance related to calculating occupational doses for monitored 58 individuals and has provided criteria regarding which individuals should be monitored for radiation 59 exposure. Such guidance includes the following:

60

• RG 8.7, "Instructions for Recording and Reporting Occ upational Radiation 61 Exposure Data" (Ref. 4) 62 63

• RG 8.9, Revision 1, "Acceptable Concepts, Models, Equations, and Assumptions 64 for a Bioassay Program" (Ref. 5) 65 66

• RG 8.11, "Applications of Bioassay for Uranium" (Ref. 6) 67 68

• RG 8.25, Revision 1, "Air Sampling in the Workplace" (Ref. 7) 69 70

• RG 8.29, "Instruction Concerning Risks from Occupational Radiation Exposure" 71 (Ref. 8) 72 73 RG 8.34, Revision 1, Page 3

• RG 8.35, Revision 1, "Planned Special Exposures" (Ref. 9) 74 75

• RG 8.36, "Radiation Dose to the Embryo/Fetus" (Ref. 10) 76 77

• RG 8.40, "Methods for Measuring Effective Dose Equivalent from External 78 Exposure" (Ref. 11) 79 80 81 Purpose of Regulatory Guides 82 The NRC issues RGs to describe to the public methods that the staff considers acceptable 83 for use in implementing specific parts of the agency's regulations, to explain techniques that the 84 staff uses in evaluating specific problems or postulated accidents, and to provide guidance to 85 applicants. RGs are not substitutes for regulations and compliance with them is not required.

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

89 Paperwork Reduction Act 90 This RG discusses information-collection requirements covered by 10 CFR Part 20 and 91 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," that the Office of 92 Management and Budget (OMB) approved under OMB control numbers 3150

-0014 93 and 3150-0011 respectively. The NRC may neither conduct nor sponsor, and a person is not 94 required to respond to, an information-collecti on request or requirement unless the requesting 95 document displays a currently valid OMB control number.

96 97 RG 8.34, Revision 1, Page 4 TABLE OF CONTENTS 98 A. INTRODUCTION ...............................................................................................................

...... 1 99 P URPOSE ..............................................................................................................................

......... 1 100 APPLICABLE RULES AND R EGULATIONS

...................................................................................... 1 101 R ELATED G UIDANCE .................................................................................................................... 2 102 PURPOSE OF R EGULATORY G UIDES ............................................................................................. 3 103 PAPERWORK R EDUCTION A CT ..................................................................................................... 3 104 B. DISCUSSION .................................................................................................................

........... 5 105 REASON FOR R EVISION ................................................................................................................ 5 106 BACKGROUND .............................................................................................................................. 5 107 OCCUPATIONAL D OSE L IMITS FOR A DULTS , M INORS , AND E MBRYOS/FETUSES ........................ 6 108 P LANNED SPECIAL EXPOSURES (PSE S) ....................................................................................... 6 109 S URVEYS ..............................................................................................................................

........ 6 110 MONITORING AT LEVELS SUFFICIENT T O DEMONSTRATE C OMPLIANCE .................................... 7 111 USE OF EFFECTIVE DAC S ............................................................................................................ 7 112 A LPHA MONITORING AT N UCLEAR P OWER P LANTS ................................................................... 7 113 DISCRETE R ADIOACTIVE-P ARTICLE MONITORING AND SDE ...................................................... 8 114 HARMONIZATION WITH I NTERNATIONAL S TANDARDS ................................................................ 9 115 D OCUMENTS D ISCUSSED IN STAFF R EGULATORY GUIDANCE .................................................... 9 116 C. STAFF REGULATORY GUIDANCE .................................................................................. 10 117 1. MONITORING C RITERIA ..................................................................................................... 10 118 2. OCCUPATIONAL D OSE ........................................................................................................ 11 119 3. P ROSPECTIVE ASSESSMENTS OF THE NEED FOR OCCUPATIONAL DOSE M ONITORING ..... 11 120 4. DETERMINATION OF E XTERNAL D OSES ............................................................................. 12 121 5. DETERMINATION OF INTAKES ............................................................................................ 14 122 6. DETERMINATION OF I NTERNAL D OSES .............................................................................. 15 123 7. USE OF INDIVIDUAL OR M ATERIAL-SPECIFIC INFORMATION ............................................ 18 124 8. L IMITATION ON URANIUM I NTAKE .................................................................................... 18 125 9. R ECORDING O F INDIVIDUAL M ONITORING RESULTS ........................................................ 18 126 D. IMPLEMENTATION ............................................................................................................

19 127 REFERENCES ....................................................................................................................

......... 20 128 APPENDIX A, METHODS OF CALCULATING INTERNAL DOSE ................................. 23 129 130 131 132 133 134 135 RG 8.34, Revision 1, Page 5 B. DISCUSSION 136 137 Reason for Revision 138 This revision of RG 8.34 provides updated regulatory guidance on monitoring criteria and 139 methods of calculating occupational dose based on the revised definition of the TEDE. This RG 140 also provides updated guidance on acceptable methods of:

141

• Determining the need for monitoring and demonstrating complianc e with occupational 142 dose limits.

143

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

144

• Assessing deep

-dose equivalent (DDE) when the measurements of the primary monitoring 145 device (dosimeter) are inconsistent with other radiological measurements (e.g., surveys or 146 electronic dosimeters).

147

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

148

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

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

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

-uniform conditions, the 167 previous TEDE definition tended to provide dose assessments that were excessively conservative.

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

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

173 RG 8.34, Revision 1, Page 6 174 Occupational Dose Limits for Adults, Minors, and Embryos/Fetuses 175 For adults, occupational dose limits (except for planned special exposures) are established in 176 10 CFR 20.1201(a) as follows:

177

• For protection against stochastic effects, the annual TEDE limit is 5 rem 178 (50 millisieverts (mSv)).

179

• For protection against nonstochastic effects, the annual total organ dose equivalent 180 (TODE) limit is 50 rem (500 mSv).

181

• For protection of the lens of the eye, the annual lens dose equivalent (LDE) limit is 182 15 rem (150 mSv).

183

• For protection of the skin of the whole body or of the skin of any extremity, the 184 annual shallow

-dose equivalent (SDE) limit is 50 rem (500 mSv).

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

187 For the embryo/fetus of a declared pregnant woman, a dose equivalent limit during the entire 188 pregnancy is established in 10 CFR 20.1208, "Dose Equivalent to an Embryo/Fetus," as 0.5 rem 189 (5 mSv). 190 Planned Special Exposures (PSEs) 191 PSEs are subject to the conditions specified in 10 CFR 20.1206, "Planned Special 192 Exposures" (e.g., exceptional circumstances, specific authorizations, and informing and instructing 193 the worker). RG 8.35, "Planned Special Exposures," provides guidance on conducting PSEs. For 194 dose-accounting purposes, dose received during a PSE is in addition to and accounted for 195 separately from the dose that is limited by 10 CFR 20.1201.

196 Surveys 1 197 Surveys (i.e., evaluations of the radiological conditions and potential hazards) should be 198 conducted as necessary in support of radiological monitoring and calculation of occupational dose.

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

201 When a licensee assigns or permits the use of respiratory protection equipment to limit the 202 intake of radioactive material, 10 CFR 20.1703(c)(2) requires surveys and bioassays, as necessary, 203 to evaluate actual intakes. Indications of an intake could include facial contamination, nasal 204 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 7 contamination, malfunctioning respiratory protection equipment, loss of engineering controls 205 creating an airborne radioactivity area, and work in unknown or unplanned airborne radioactivity 206 areas. 207 During operations, licensees should perform airborne radioactivity surveys as required in 208 10 CFR 20.1502 to characterize the radiological hazards that may be present and, as appropriate, 209 use engineering and respiratory protection equipment to reduce intakes. When it is not practical to 210 use process or engineering controls to reduce the concentrations of airborne radioactivity to values 211 below those that define an airborne radioactivity area, licensees are required under 212 10 CFR 20.1702(a), to be consistent with keeping the TEDE ALARA, to increase monitoring 213 (e.g., perform air sampling and track Derived Air Concentration (DAC)

-hours and bioassay 214 measurements) and to limit intakes by using access controls, limiting exposure times, or having 215 individuals use respiratory protection equipment.

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

226 Use of Effective DACs 227 The regulation at 10 CFR 20.1204(e) provides a method for determining internal exposure 228 when the identity and concentration of each radionuclide in a mixture is known. The identities and 229 concentrations of radionuclides may be determined based on representative radiological surveys 230 identifying the specific radionuclides and quantifying their relative mix. Once the relative mix is 231 known, licensees may apply scaling factors applicable to the mixture for use in calculating DACs and 232 tracking DAC

-hours as specified in 20.1204(e). This is commonly referred to as "effective DACs" 233 and is applicable to beta/gamma activity, alpha activity, and hard

-to-detect radionuclides.

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

241 242 Alpha Monitoring at Nuclear Power Plants 243 For reactor facilities that have experienced si gnificant fuel defects, alpha contamination 244 may be a radiological hazard requiring specific evaluation. Alpha contamination (when present) 245 RG 8.34, Revision 1, Page 8 requires specific evaluation because the DAC values for alpha emitting isotopes are generally 246 orders of magnitude more restrictive than DACs for beta-emitting and gamma-emitting isotopes.

247 Each facility should characterize and update its alpha source term as needed based on the 248 facility's operational history. Alpha source

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

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

260 The principal transuranic nuclides producing alpha radiological hazards include the 261 isotopes of curium, plutonium, and americium. For historical fuel failures (e.g., ten years have 262 passed since significant fuel failure), the shorter

-lived curium

-242 will have largely decayed, 263 leaving the longer

-lived alpha radionuclides with more restrictive DACs and annual limits on 264 intake (ALI) as the most prevalent hazard. However, investigations of more recent fuel failures are 265 likely to identify curium

-242 as the most abundant alpha-emitting nuclide, which has less 266 restrictive DAC and ALI values. Therefore, effective DAC values must be updated as needed to 267 account for the time

-dependent (decayed) mix of alpha radi onuclides. In addition, consideration 268 should be given to transuranic isotopes which decay by other than alpha emission (e.g., Pu-241).

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

271

• knowledge of the specific alpha radionuclide mix 272

• knowledge of the solubility/insolubility of the radionuclides 273

• conservative assumptions about the most restrictive radionuclide in the mixture 274

• determination of site-specific effective-DAC alpha values 275 Discrete Radioactive-Particle Monitoring and SDE 276 A discrete radioactive particle (DRP) is a small (usually microscopic) and highly 277 radioactive particle emitting either only beta or both beta a nd gamma radiation and having 278 relatively high specific activity. DRPs are primarily an external exposure hazard to the skin, as 279 measured by the SDE.

280 In 2002, the NRC amended its regulations related to the shallow-dose equivalent/skin-dose 281 limit in 10 CFR Part 20 (at 67 FR 16298 (Ref. 13); see also Regulatory Issue Summary 2002

-10, 282 "Revision of the Skin Dose Limit in 10 CFR Part 20" (Ref. 14)). The amended regulations 283 changed the definition and method of calculating SDEs by specifying that the assigned SDE must 284 be the dose averaged over the contiguous 10 cm 2 of skin receiving the highest exposure.

285 RG 8.34, Revision 1, Page 9 Harmonization with International Standards 286 The NRC has a goal of harmonizing its guidance (to the extent that this is practical) with 287 international standards. The International Commission on Radiological Protection (ICRP) and the 288 International Atomic Energy Agency (IAEA) have issued a significant number of standards, 289 guidance and technical documents, and recommendations addressing good practices in most 290 aspects of radiation protection. The NRC encourages licensees to consult the international 291 documents noted throughout this guide and implement the applicable good practices they contain 292 that are consistent with NRC regulations.

293 Such documents include the following:

294

• ICRP Publication 26, "Recommendations of the International Commission on 295 Radiological Protection" (Ref. 15) 296

• ICRP Publication 30, (7-volume set including supplements), "Limits for Intakes of 297 Radionuclides by Workers" (Ref. 16) 298

• ICRP Publication 54, "Individual Monitoring for Intakes of Radionuclides by Workers" 299 (Ref. 17) 300

• ICRP Publication 60, "1990 Recommendations of the International Commission on 301 Radiological Protection" (Ref. 18) 302

• ICRP Publication 68, "Dose Coefficients for Intakes of Radionuclides for Workers" 303 (Ref. 19) 304

• ICRP Publication 78, "Individual Monitoring for Internal Exposure of Workers" (Ref. 20) 305

• ICRP Publication 103, "The 2007 Recommendations of the International Commission on 306 Radiological Protection" (Ref. 21) 307 Documents Discussed in Staff Regulatory Guidance 308 Although this RG uses information, in part, from one or more reports developed by 309 external organizations and other third-party guidance documents, the RG does not endorse these 310 references other than as specified in this RG. These reports and third

-party guidance documents 311 may contain references to other reports or third

-party guidance documents ("secondary 312 references"). If a secondary reference has itself been incorporated by reference in NRC regulations 313 as a requirement, licensees and applicants must comply with that requirement in the regulation.

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

323 324 RG 8.34, Revision 1, Page 10 C. STAFF REGULATORY GUIDANCE 325 326 1. Monitoring Criteria 327 328 Regulations in 10 CFR 20.1502 require individual monitoring of external and internal 329 occupational dose at levels sufficient 2 to demonstrate compliance with the occupational dose 330 limits. As a minimum, licensees must monitor occupational exposure to radiation from licensed and 331 unlicensed radiation sources 3 under the control of the licensee.

332 333 For external occupational exposure, licensees are required to supply and require the use of 334 individual monitoring devices if the external occupational dose:

335 336

• for adults, is likely to exceed 10 percent of the occupational dose limits in 337 10 CFR 20.1201(a);

338

• for minors, in one year, is likely to exceed a deep-dose equivalent of 0.1 rem (1 mSv), 339 a lens dose equivalent of 0.15 rem (1.5 mSv), or a shallow-dose equivalent to the skin 340 of the whole body or to the skin of the extremities of 0.5 rem (5 mSv); or 341

• for declared pregnant women, during their entire pregnancy, is likely to exceed a 342 deep-dose equivalent of 0.1 rem (1 mSv), and 343

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

344 345 For internal occupational exposure, licensees are required to monitor the intake of 346 radioactive material and assess the CEDE by 10 CFR 20.1502(b) if the intake is likely to exceed:

347

• 10 percent of the applicable annual limit on intake (ALI) for adults; 348

• 0.1 rem (1 mSv) for minors in one year; or 349

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

350 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).

RG 8.34, Revision 1, Page 11

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

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

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

362 3. Prospective Assessments of the Need for Occupational Dose Monitoring 363 Licensees must identify those individuals recei ving occupational dose, either individually 364 or as a group or category of individuals. Individuals pre

-designated by the licensee as receiving 365 occupational dose are subject to the occupational dose limits; otherwise, individuals must be 366 considered as members of the public subject to public dose limits in 10 CFR 20.1301, "Dose Limits 367 for Individual Members of the Public."

368 Once occupationally exposed individuals are identified, licensees should perform a 369 prospective assessment to determine whether those individuals are "likely to exceed" the minimum 370 exposure levels specified in 10 CFR 20.1502 (i.e., to determine the need for monitoring of the 371 occupational dose). The potential for unlikely exposures and accident conditions need not be 372 considered because these events, by definition, are unlikely. However, as discussed at 373 60 FR 36039, the term "likely to receive" includes "normal situations as well as abnormal 374 situations involving exposure to radiation which can reasonably be expected to occur during the life 375 of the facility." Therefore, licensees should consider normal operations and anticipated operational 376 occurrences (e.g., unplanned onsite events, such as sudden increases in external radiation levels, or 377 localized areas of high airborne radioactivity) but would not need to consider design

-basis 378 accidents 379 The prospective assessment determines the type of monitoring required (e.g., external-dose or 380 internal-dose monitoring). In performing a prospective assessment, an evaluation should be 381 performed based on planned work activities and likely exposure conditions. In the prospective 382 assessment, licensees may take credit for the use of engineering controls (e.g., containment, 383 decontamination, ventilation, and filtration). However, if licensees are using respiratory protection 384 equipment to limit the intake of radioactive material, licensees must establishing a respiratory 385 protection program and perform air sampling, surveys, and bioassays to evaluate intakes and 386 estimate dose in accordance with the 10 CFR 20.17

03. Prospective assessments should be revised 387 when there are substantial ch anges to the radiological conditions of personnel exposure 388 (e.g., changes in work activities, airborne c oncentrations, beta energy spectra, or use of 389 radiation-producing equipment emitting new or different types of energies).

390 391 The requirements for monitoring in 10 CFR 20.1502 refer to exposures that might occur at 392 each licensee individually. Doses that have already been received while in the employ of another 393 licensee, or that might be received in the future while in the employ of another licensee or 394 unlicensed entity, are excluded from consideration in a licensee's determination of the need to 395 RG 8.34, Revision 1, Page 12 monitor an individual. The need for monitoring should be based on the anticipated exposure to 396 licensed or unlicensed sources under the control of a single licensee.

397 4. Determination of External Doses 398 a. Determination of the TEDE 399 Under 10 CFR 20.1202, if a licensee is required to monitor both external dose and internal 400 dose, the licensee must demonstrate compliance with the dose limits by summing external and 401 internal doses (i.e., TEDE = EDEX + CEDE). However, if the licensee is required to monitor only 402 external doses under 10 CFR 20.1502(a) or only internal doses under 10 CFR 20.1502(b), 403 summation is not required to demonstrate compliance with the occupational dose limits. For 404 example, if the internal dose is not monitored, the CEDE can be assumed to be equal to zero and the 405 TEDE is equal to the EDEX. Similarly, if the external dose is not monitored, the EDEX can be 406 assumed to be equal to zero and the TEDE is equal to the CEDE.

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

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

413 2. Measuring external exposure with one or more external personal monitoring devices and 414 determining EDEX using an NRC

-approved method (such as those provided in RG 8.40 or 415 as specifically approved elsewhere by the NRC).

416 3. Calculating the EDEX based on survey data obtained under 10 CFR 20.1501 or on other 417 radiological data (such as known source activity, dose rates, and exposure times) using 418 scientifically sound technical methods. This might be required (a) under unique exposure 419 situations (e.g., if an individual's body were partially exposed to radiation streaming in a 420 narrow beam geometry), (b) when the individual's monitoring device was not in the region 421 of the highest whole

-body exposure (in accordance with 10 CFR 20.1201(c)), or (c) when 422 the results of the individual monitoring are not available (i.e., the monitoring device is 423 damaged or lost).

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

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

429 c. Determination of the Deep-Dose Equivalent (DDE) 430 The DDE (external exposure of the whole body) is typically measured with a passive 431 primary monitoring device that assesses the dose at a tissue depth of 1 centimeter (cm) (a mass 432 thickness of 1,000 mg/cm 2). The DDE can also be calculated if the appropriate parameters are 433 known (i.e., the radiation source strength, the exposure geometry, and whether full or partial 434 shielding was in place).

435 RG 8.34, Revision 1, Page 13 An individual monitoring device located at the most highly exposed part of the whole body 436 measuring the DDE is a conservative and (for uniform exposures) a reasonably accurate estimate of 437 the EDEX. However, if the radiation dose is highly non uniform, causing a specific part of the 438 whole body (head, trunk, arms above the elbow, or le gs above the knees) to receive a substantially 439 higher dose than the rest of the whole body, the individual monitoring device should be placed near 440 that part of the whole body expected to receive the highest dose. There are several other 441 NRC-approved methods for determining EDEX provided in RG 8.40.

442 443 In many exposure situations, a required monitoring device (e.g., a passive dosimeter) may 444 be voluntarily supplemented with an additional, active dosimeter (e.g., an electronic dosimeter 445 used for work control and daily dose accounting pu rposes). Due to the differences in dosimeter 446 design and detection technology, and the relative measurement errors associated with each type of 447 dosimeter, there can be valid differences in readings of these two dosimeters for the same exposure, 448 even if the dosimeters are co-located on the monitored individual. Within a reasonable, licensee 449 pre-determined accuracy criteria (depending on dosimeter designs), small differences between 450 measurements can be disregarded and either dosimetry value used as the measured dose (since both 451 results are considered valid and equal within measurement error). However, a significantly higher 452 reading on the voluntary dosimeter may indicate that the required dosimeter was not appropriately 453 placed to measure the highest exposed part of the whole body. Licensees should investigate those 454 cases where a significant discrepancy exists between dosimeters. If the differences cannot be 455 resolved, an assessment must be performed to determine the DDE, LDE, and SDE for the highest 456 exposed part of the whole body, as provided for in 10 CFR 20.1201(c).

457 d. Determining the LDE 458 If the LDE is being monitored with a dosimete r, that dosimeter should be calibrated to 459 measure the dose at a tissue depth of 0.3 centimeter (cm) (a mass thickness 300 mg/cm 2). 460 Alternatively, the LDE may be conservatively determined based on SDE measurements at 461 7 mg/cm 2. In many exposure situations, safety glasses can be worn to minimize exposures to the 462 lens of the eye from low

-energy (or poorly penetrating) radiations, potentially eliminating the need 463 for monitoring the LDE.

464 e. Determination of the SDE 465 The SDE is defined only for external exposure at a tissue depth of 0.007 cm (a mass 466 thickness of 7 mg/cm 2), and is the dose averaged over the contiguous 10 cm 2 of skin receiving the 467 highest exposure. If the SDE is being measured with a dosimeter, that dosimeter should be 468 calibrated to measure the dose at a tissue depth of 7 mg/cm

2. For skin contamination, the computer 469 code described in NUREG/CR

-6918, "VARSKIN: A Computer Code for Skin Contamination 470 Dosimetry" (Ref. 23) may be used to assess the SDE. The SDE may also be determined from 471 analytical calculational methods based on survey data when dosimetry methods are not 472 representative of the actual exposure conditions.

473 The SDE for exposure to submersion-class radionuclides containing low

-energy betas is 474 not readily measurable by direct survey techniques or dosimetry methods and hence may need to be 475 calculated based on air-sample analyses and DAC

-hr tracking. This submersion exposure 476 information may be needed for informing workers of radiological exposure conditions 477 (e.g., informing workers of the SDE rates during pre-job briefings) and also to account in dose 478 records for the SDE that might not be adequately measured by dosimeters (e.g., because of the 479 dosimeter's lack of response to a low

-energy beta spectrum).

480 RG 8.34, Revision 1, Page 14

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

486 The assessment of intake should include not only the readily detected radionuclides but 487 also the hard

-to-detect radionuclides if their dose contribution is significant. The activity of 488 hard-to-detect radionuclides may be based on scaling factors that correspond to the amount of 489 readily detected radionuclides. See RG 8.25, "Air Sampling in the Workplace," and Regulatory 490 Guide 8.9, "Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program,"

491 for further guidance on determining uptakes and intakes.

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

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

497 If respiratory protection is used to limit the intake of radioactive materials, 498 10 CFR 20.1703(c)(4)(i) requires internal monitoring to be implemented as part of the respiratory 499 protection program. When respiratory protection is provided, the intake is adjusted by dividing the 500 ambient air concentration by the appropriate Assigned Protection Factor (APF) listed in 501 Appendix A, "Assigned Protection Factors for Respirators," to 10 CFR 20. If the ambient air 502 concentration is determined by performing breathing

-zone air sampling inside the respiratory 503 protective device (such as with a lapel air sampler inside a loose

-fitting supplied air hood or suit), 504 no APF adjustment is made to the ambient air concentration as measured in the breathing

-zone air 505 sample. 506 a. Determining the Intake Based on Air Sampling 507 Intake (I) based on air-sampling results can be assessed by multiplying the airborne 508 concentration (C) by the breathing rate and the exposure time:

509 I = CAir sample (µCi/ml)

• breathing rate (ml/minutes)

• exposure time (minutes), where the 510 breathing rate of a "Reference Man" under light working conditions is 2E+4 ml/minute 511 (20 liters/minute).

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

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

518 I = AAir sample (µCi)

• breathing rate/air sampler flow rate (ml/min), where the breathing rate 519 of a "Reference Man" under light working conditions is 2E+4 ml/minute 520 (20 liters/minute).

521 RG 8.34, Revision 1, Page 15

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

528 Any intake from wounds is generally assessed based on bioassay measurements using a 529 combination of whole body in vivo bioassay and handheld instrumentation. The bioassay 530 measurements should determine the location and depth of the injected source so that CDE dose 531 calculations may be made to the most highly exposed 10 cm 2 area of the skin at a depth of 0.007 cm 532 (see Section 6.d below).

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

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

544 Internal doses may also be assessed based on whole-body count data and scaling factors 545 when nominal (e.g., less than 500 mrem CEDE) alpha doses occur. However, when an alpha intake 546 resulting in alpha doses exceeding a nominal quantity is considered likely, excreta sampling or lung 547 counting may be needed to assess intakes and assign dose. When excreta sampling is to be 548 initiated, sampling should begin as soon as possible following detection of the exposure and should 549 continue for a 24

-hour period or until at least one sample is collected (following the first void for 550 urine). ANSI N13.39-2001 (R2011), "Design of Internal Dosimetry Programs" (Ref. 25), provides 551 additional guidance on excreta sampling.

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

557 RG 8.34, Revision 1, Page 16

• Using dose coefficients 4 from the U.S. Environmental Protection Agency's Federal 558 Guidance Report No. 11 (FGR-11) (Ref. 26).

559

• Using ALI methods.

560

• Using DAC-hour methods.

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

562 Note: When performing CEDE calculations using the ALI and DAC-hour methods, the 563 ALI and DAC values provided in Appendix B, "Annual Limits on Intake (A LIs) and Derived Air 564 Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; 565 Concentrations for Release to Sewerage," to 10 CFR Part 20 must be used unless the licensee has 566 obtained prior NRC approval in accordance with 10 CFR 20.1204(c)(2) to adjust the ALI or DAC 567 values. 568 b. Calculation of the Committed Dose Equivalent (CDE) 569 The CDE is the 50-year committed dose equivalent from the intake of radioactive material.

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

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

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

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

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

581 If both internal and external monitoring are being performed, the licensee must 582 demonstrate that both the 5-rem TEDE and the 50-rem TODE limits are met. One method of 583 demonstrating compliance with the TODE limit is by summing the DDE and the CDE to the most 584 highly exposed organ. Another acceptable method of demonstrating that the TODE limit is met is 585 by keeping the maximum DDE below 5 rem and the CEDE below 1 rem 5; if this is done, the TODE 586 cannot exceed its 50-rem limit. In this case, the CDE does not need to be determined because 587 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 17 compliance was demonstrated by calculation. If the CEDE does exceed 1 rem, the CDE must be 588 determined in order to demonstrate compliance with the dose limits.

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

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

593

• TODE limits (10 CFR 20.1201(a)(1)(ii))-the sum of the DDE and the CDE to any 594 individual organ or tissue other than the lens of the eye being equal to 50 rem 595 (0.5 Sv)-and 596

• SDE limits (10 CFR 20.1201(a)(2)(ii))-the SDE to the skin of the whole body or skin of 597 any extremity being equal to 50 rem (0.5 Sv).

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

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

604 In making the TODE dose calculation (to the skin organ) under 20.1201(a)(1)(ii), the DDE 605 component is zero (because DDE is specifically defined as an external whole-body exposure). As a 606 result, the CDE is determined for the basal layer of the skin at a depth of 0.007 cm below skin 607 surface for the most highly exposed, contiguous 10-cm 2 area. 608 In summary, the CDE to the skin is the appropriate quantity to be calculated as the 609 integrated dose from the time of injection to the time the source is removed or by the 50-year 610 integration period for committed dose. The CDE is to be determined at a depth of 611 0.007 centimeters below the surface of the skin, averaged over the most highly exposed 10 cm 2 of 612 the basal layer of the skin. In order to do this calculation, the location (depth) of the source and 613 distance to the basal layer must be determined as an input parameter. The VARSKIN computer 614 code may be used in performi ng the CDE skin-dose calculations.

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

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

627 e. Calculating the CDE and CEDE for Inhalation, Submersion and Absorption 628 RG 8.34, Revision 1, Page 18 A number of methods are acceptable for calculating the CDE and CEDE from the intake of 629 radioactive materials. Some of these methods are described below. However, calculations of the 630 CEDE must be based on organ weighting factors and tissues specified in 10 CFR Part 20. The dose 631 coefficients based on ICRP Publication 60 cannot be used unless specifically approved by the 632 NRC, because ICRP 60 and ICRP 103 tissues and weighting factors are different from those in 633 10 CFR Part 20.

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

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

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

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

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

661 662 Licensees should avoid entering doses on NRC Form 5 with more significant figures than justified 663 by the precision of the basic measured values. In general, it is appropriate to enter dose values with 664 two significant figures on NRC Form 5 using the standard rules for round-off. Thus, a 665 computer-generated calculated dose of "1.726931 rems" should be entered on NRC Form 5 as "1.7 666 rems." However, licensees should generally carry at least three significant figures in calculations to 667 avoid loss of accuracy due to multiple round-offs.

668 669 In addition, licensees should not enter doses smaller than 0.001 rem on NRC Form 5 because 670 smaller values are insignificant relative to the dose limits. Therefore, a calculated committed 671 RG 8.34, Revision 1, Page 19 effective dose equivalent of "0.006192 rem" should be entered as "0.006 rem," and a value of 672 "0.000291 rem" should be entered as "0 rem." 673 D. IMPLEMENTATION 674 675 The purpose of this section is to provide information to applicants and licensees regarding 676 the NRC's plans for using this RG.

677 Methods or solutions that differ from those described in this regulatory guide may be 678 deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that 679 the proposed alternative complies with the appropriate NRC regulations. Current licensees may 680 continue to use guidance the NRC found acceptable for complying with the identified regulations 681 as long as their current licensing basis remains unchanged. 682 683 RG 8.34, Revision 1, Page 20 REFERENCES 6 684 685 1. U.S. Code of Federal Regulations (CFR), "Standards for Protection against Radiation,"

686 Part 20, Chapter I, Title 10, "Energy."

687 2. 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," Part 50, 688 Chapter I, Title 10, "Energy."

689 3. 10 CFR 19, "Notices, Instructions, and Reports to Workers: Inspection and 690 Investigations," Part 19, Ch apter I, Title 10, "Energy."

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

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

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

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

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

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

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

707 11. NRC, "Methods for Measuring Effective Dose Equivalent from External Exposure,"

708 RG 8.40, July 2010, ADAMS Accession No. ML100610534.

709 12. NRC, "Occupational Dose Records, Labeling Containers, and the Total Effective Dose 710 Equivalent,"

Federal Register, Vol. 72, No. 232, December 4, 2007, pp. 68043-68059 711 (72 FR 68043).

7 712 6 Publicly available NRC published documents are available electronically through the NRC Library on the NRC's public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRC's 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 NRC's 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 21

13. NRC, "Revision of the Skin Dose Limit,"

Federal Register, Vol. 67, No. 66, April 5, 2002, 713 pp. 16298-16301 (67 FR 16298).

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

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

718 Pergamon Press, 1977.

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

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

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

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

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

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

731 22. NRC, "Radiation Protection Requirements: Amended Definitions and Criteria,"

Federal 732 Register, Vol. 60, No. 134, July 13, 1995, pp. 36038-36043 (60 FR 36038).

733 23. NRC, "VARSKIN 5: A Computer Code for Skin Contamination Dosimetry,"

734 NUREG/CR-6918, Rev. 2, July 2014, Accession No. ML14204A361.

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

737 25. American National Standards Institute (ANSI), "Design of Internal Dosimetry Programs,"

738 ANSI N13.39-2001 (R2011), Washington, DC, 2011.

739 26. Eckerman, K.F., A.B. Wolbarst, and A.C.B. Richardson, "Limiting Values of 740 Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, 741 Submersion, and Ingestion," Federal Guidance Report No. 11 (EPA 520/1-8-020), 742 U.S. Environmental Protection Agency, Washington, DC, 1988.

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

746 28. Toohey, R.E., et al., "Dose Coefficients for Intakes of Radionuclides via Contaminated 747 Wounds," Health Physics 100(5):508-14, May 2011; a much larger and revised version 748 (Ver. 2, August 2014) is available from the Oak Ridge Institute for Science and Education 749 RG 8.34, Revision 1, Page 22 at http://orise.orau.gov/reacts/resources/retention-intake-publication.aspx (accessed 750 October 10, 2014).

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

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

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

759 ML14322A420.

760 761 762 RG 8.34, Revision 1, Page 23 Appendix A 763 Methods of Calculating Internal Dose 764 765 766 1. Calculations of the CDE and the CEDE Based on Bioassay Measurements 767 Using Federal Guidance Report No. 11 (FGR-11) 768 This method is based on using tabulated dose co efficients to calculate the dose. FGR-11 769 provides tables of dose coefficients (DCs) (FGR-11 uses the terminology "dose conversion factors")

770 for intakes by inhalation and by ingestion (see excerpt below for inhalation of cobalt-60 (Co-60)).

771 FGR-11 provides two types of DCs:

772 773 (1) DCs for the CDE to an organ or tissue per unit of activity (DCorgan) (e.g., the 774 heading "Lung" below) and 775 776 (2) DCs for the CEDE per unit of activity (DC effective) (as shown in the far right column 777 of the tables under the heading "Effective").

778 779 If site-specific information is known about the type of compound and its clearance class, the 780 appropriate clearance class can be selected. If not, the class is normally selected based on the most 781 conservative class; in Example 1, the DC for the lung is selected from clearance Class Y, which has a 782 value of 3.45E-7). Multiplying the DCs by the intake (I) for that radionuclide yields the CDE and 783 CEDE for that radionuclide.

784 785 CDE (rem) = DCorgan (rem/µCi [rem per millicurie])

• I (µCi) 786 CEDE (rem) = DC effective (rem/Ci)

• I (Ci) 787 788 Example 1: Calculations of the CDE and the CEDE for Co-60, based on bioassay 789 measurements using the DCs from FGR-11. Note: The DCs in FGR-11 are tabulated in Sieverts 790 per Becquerel (Sv/Bq) and may be converted to millirem per microcurie (mrem/Ci) by 791 multiplying by 3.7E+9.

792 793 RG 8.34, Revision 1, Page 24 794 An intake by inhalation was estimated by a whole body count to be 360 nanocuries (nCi) 795 (0.36 µCi) of Co-60 as a Class Y aerosol. Calculate the CDE to the lung and the CEDE.

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

798 799 800 DClung = (3.45E-7 Sv/Bq) * (3.7E+9) = 1277 mrem/µCi 801 DC effective = (5.91E-8 Sv/Bq) * (3.7E+9) = 219 mrem/µCi 802 803 The doses are calculated by multiplying these DCs by the intake of 0.36 µCi:

804 805 CDElung = (1277 mrem/µCi) * (0.36 µCi) = 460 mrem 806 CEDE = (219 mrem/µCi) * (0.36 µCi) = 79 mrem 807 808 2. Calculation of the CEDE based on Bi oassay Measurements using Stochastic 809 ALIs 810 The ALI values are listed in Table 1 of 10 CFR 20, Appendix B, "Annual Limits on Intake 811 (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; 812 Effluent Concentrations for Release to Sewerage." Column 1 lists the values for oral ingestion and 813 Column 2 lists the values for inhalation. The stochastic ALI values can be used in the calculation 814 of the CEDE, which is based on the fraction of the allowable annual intake and the 5-rem 815 (50-millisievert (mSv)) CEDE dose limit. When the ALI is defined by the stochastic limit, this 816 value alone is given in the table.

817 818 819 820 RG 8.34, Revision 1, Page 25 Because the stochastic ALI corresponds to a 5-rem (50-mSv) CEDE dose limit, the CEDE 821 may be calculated based on the ratio of the intake to the stochastic ALI multiplied by 5 rem 822 (50 mSv):

823 824 CEDE = (I/ALI)

• 5 rem 825 826 Example 2: Calculate the CEDE based on bioassay measurements using the stochastic ALI.

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

829 830 From Appendix B above, Table 1, Column 2, the ALI for Class Y Co-60 is:

831 832 ALI (stochastic) = 30 µCi 833 CEDE = (I/ALI)

• 5 rem 834 CEDE = (0.36 µCi/30 µCi)

• 5 rem = 0.06 rem = 60 mrem 835 836 Note: Doses calculated based on FGR-11 methods are generally more precise than doses 837 calculated based on ALI values, because ALI values are given to only one significant figure.

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

843 844 For Co-60, a 60-mrem value based on an ALI calculation compares to a calculated CEDE 845 value of 79 mrem using the FGR-11 method as determined in Example 1 above. For other 846 radionuclides such as Co-58, the differences might be larger. However, either calculational method 847 and/or result is acceptable in demonstrating compliance with regulatory limits.

848 849 3. Calculation of the CDE Based on Bioassay Measurements Using 850 Nonstochastic ALI 851 The 10 CFR 20 Appendix B , Table 1, Column 2, nonstochastic ALI values can be used in 852 the calculation of the CDE, based on the fraction of the allowable annual intake and the 50-rem 853 (500-mSv) CDE dose limit. When the ALI is defined by the nonstochastic limit, this value is listed 854 first in the table with its corresponding organ (see excerpt below), and the corresponding stochastic 855 ALI are given in parentheses (e.g., 9E+1 µCi (90 µCi) for ingestion and 2E+2 µCi (200 µCi) for 856 inhalation in the excerpt below).

857 RG 8.34, Revision 1, Page 26 858 859 860 Because the nonstochastic ALI corresponds to a 50-rem (500-mSv) CDE dose limit, the 861 CDE may be calculated based on the ratio of the intake to the nonstochastic ALI multiplied by 862 50 rem (500 mSv):

863 CDE = (I/ALI)

• 50 rem 864 865 Note: For a mixture of radionuclides, the "sum of the fractions" technique as described in 866 10 CFR 20.1202(b) must be used.

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

868 869 The intake by inhalation for a worker was estimated by bioassay to be 131 nCi (0.131 µCi) 870 of iodine-131 (I-131) as a Class D aerosol. Calculate the CDE to the thyroid.

871 872 From Appendix B above, Table 1, Column 2, the ALI for Class D I-131 is:

873 874 ALI (nonstochastic) = 5E+1 µCi = 50 µCi 875 CDE = (0.131 µCi/50 µCi)

• 50 rem = 0.131 rem = 131 mrem 876 877 4. Calculation of the CDE Based on Air Sampling and Nonstochastic 878 DAC-Hours (DAC-hr) 879 For nonstochastic radionuclides, an exposure to an airborne concentration of 1 DAC for 880 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.

881 CDE = (25 mrem per DAC-hr)

• number of DAC-hr 882 where the number of DAC-hr = (air concentration / DAC value)

• exposure time.

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

884 885 Calculate the CDE to the thyroid for a 30-minute exposure based on an air-sample result of 886 2.1E-7 µCi/ml from I-131.

887 RG 8.34, Revision 1, Page 27 888 The nonstochastic DAC for I-131 is listed in Appendix B (see the excerpt below) as 889 2E-8 µCi/ml.

890 891 892 CDE = 25 mrem/DAC-hr * (2.1E-7 µCi/ml / 2E-8 µCi/ml) number of DACs * (0.5 hr) =

893 131 mrem 894 5. Calculations of the CEDE Based on Air Sampling and Stochastic DAC-hr 895 For stochastic radionuclides (e.g., Co-60), an exposure to an airborne concentration of 896 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 897 a 2.5-mrem CEDE per stochastic DAC-hr.

898 CEDE = 2.5 mrem/DAC-hr

• number of DAC-hr 899 where the number of DAC-hr = (air concentration / DAC value)

• exposure time.

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

901 902 Calculate the CEDE for a 30-minute exposure based on an air sample result of 903 2.1E-7 µCi/ml from Co-60.

904 905 From Appendix B below, the stochastic DAC for Co-60 in a clearance Class Y compound 906 is 1E-8 µCi/ml.

907 908 909 RG 8.34, Revision 1, Page 28 CEDE = (2.5 mrem/DAC-hr) * [(2.1E-7 µCi/ml) / (1E-8 µCi/ml)] number of DACs

• 910 (0.5 hr) = 26 mrem 911 6. Calculation of the CEDE Based on Air Sampling and Calculated Stochastic 912 DAC-hr 913 CEDE = 2.5 mrem/DAC-hr

• number of DAC-hr 914 Number of DAC-hr = air concentration / calculated DAC value

• exposure time 915 Note: Appendix B to 10 CFR Part 20 does not list the stochastic DAC values (as shown in 916 the empty circled cell below) for radionuclides with intakes that have nonstochastic limits.

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

920 921 Example 6: Calculate the CEDE based on air sampling and calculated stochastic DAC-hr.

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

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

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

928 CEDE = (2.5 mrem/hr/DAC-hr) * [(2.1E-7 µCi/ml) / (8E-8 µCi/ml)] DACs * (0.5 hr) 929 = 3.3 mrem 930