ML14133A641

Draft Regulatory Guide 8.34, Revision 1 (Draft Was Issued as DG-8031) Monitoring Criteria and Methods to Calculate Occupational Radiation Doses

ML14133A641
Person / Time
Issue date:	05/28/2014
From:	Steven Garry NRC/NRR/DRA
To:
Garry S
References
DG-8031 RG-8.034
Download: ML14133A641 (23)

Text

U.S. NUCLEAR REGULATORY COMMISSION May 2014 OFFICE OF NUCLEAR REGULATORY RESEARCH Revision 1 REGULATORY GUIDE Technical Lead S. Garry 1 DRAFT REGULATORY GUIDE 8.34 2 (Draft was issued as DG-8031) 3 4 MONITORING CRITERIA AND METHODS TO CALCULATE 5 OCCUPATIONAL RADIATION DOSES 6 A. INTRODUCTION 7 Purpose 8 This guide provides methods acceptable to the staff of the U.S. Nuclear Regulatory Commission 9 (NRC) for monitoring the occupational radiation dose to individuals and for calculating occupational 10 radiation doses. The Regulatory Guide applies to both reactor and materials licensees under both NRC and 11 Agreement State licenses.

12 Applicable Rules and Regulations 13 The regulations established by the NRC in Title 10, Part 20, of the Code of Federal Regulations (10 14 CFR Part 20), Standards for Protection against Radiation, (Ref. 1), Section 20.1101, Radiation Protection 15 Programs, establishes requirements for licensees to limit radiation exposures to individuals within the 16 specified regulatory radiation dose limits and are as low as is reasonably achievable (ALARA). To 17 demonstrate compliance with the dose limits, licensees must perform surveys and, when appropriate, monitor 18 the radiation exposure and calculate the resultant doses.

19 Also, Section 20.1201, Occupational Dose Limits for Adults, establishes radiation dose limits for 20 occupationally exposed individuals. These limits apply to the sum of the dose received from external 21 exposure and the dose from internally deposited radioactive material. Conditions that require individual 22 monitoring of external and internal occupational doses are specified in 10 CFR 20.1502, Conditions 23 Requiring Individual Monitoring of External and Internal Occupational Dose. Monitoring the intake of 24 radioactive material and assessing the committed effective dose equivalent (CEDE) (for internal exposures) is 25 required by 10 CFR 20.1502(b). The calculations licensees are required to perform in order to comply with 26 these regulations were affected by the 2007 revision of 10 CFR Part 20, Section 20.1003, Definitions and 27 10 CFR 50, Section 50.2, Definitions, (Ref. 2). This revision redefined the Total Effective Dose Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site 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 the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/. 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. MLXXXXXXXXX. The regulatory analysis may be found in ADAMS under Accession No. MLXXXXXXXXX and the staff responses to the public comments on DG-XXXX may be found under ADAMS Accession No. MLXXXXXXXX. [Include any other document related to the RG development that may be of interest to the public]

28 Equivalent (TEDE), as the sum of the effective dose equivalent (for external exposures) and the CEDE (for 29 internal exposures).

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

31 10 CFR 20.1007, Communications, 32 10 CFR 20.1202, Compliance with Requirements for Summation of External and Internal Doses, 33 10 CFR 20.1204, Determination of Internal Exposure, 34 10 CFR 20.1206, Planned Special Exposures, 35 10 CFR 20.1207, Occupational Dose Limits for Minors, 36 10 CFR 20.1208, Dose Equivalent to an Embryo/Fetus, 37 10 CFR 20.1501 Subpart F - Surveys and Monitoring (General) 38 10 CR 20.1703, Use of Individual Respiratory Protection Equipment, 39 10 CFR 20.2106, Records of Individual Monitoring Results, 40 10 CFR 20.2206, Reports of Individual Monitoring, and 41 10 CFR Part 19, Notices, Instructions, and Reports to Workers: Inspection and Investigations, 42 (Ref. 3).

43 Related Guidance 44 The NRC has developed guidance related to calculating occupational doses for monitored individuals 45 and provided criteria regarding which individuals should be monitored for radiation exposure. Such guidance 46 includes:

47

• Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation 48 Exposure Data, U.S. Nuclear Regulatory Commission, Washington, DC.

49

• Regulatory Guide 8.9, Revision 1, Interpretation of Bioassay Measurements (Ref. 4),

50 provides methods of determining intakes from bioassay results, 51

• Regulatory Guide 8.11, Applications of Bioassay for Uranium 52

• Regulatory Guide 8.25, Revision 1, Air Sampling in the Workplace (Ref. 5), provides 53 methods of determining intakes from air sampling measurements, 54

• Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation 55 Exposure DRAFT RG 8.34, Revision 1, Page 2

56

• Regulatory Guide 8.35, Revision 1, "Planned Special Exposures (Ref. 6), provides guidance 57 on conducting planned special exposures, 58

• Regulatory Guide 8.36, Radiation Dose to the Embryo/Fetus (Ref. 7), provides methods of 59 calculating doses to the embryo/fetus, and 60

• Regulatory Guide 8.40, Methods for Measuring Effective Dose Equivalent from External 61 Exposure, (Ref. 8), provides details on acceptable methods of determining the effective 62 dose equivalent (from external exposure).

63 Purpose of Regulatory Guides 64 The NRC issues regulatory guides to describe to the public methods that the staff considers 65 acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the 66 staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants.

67 Regulatory guides are not substitutes for regulations and compliance with them is not required. Methods and 68 solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis 69 for the findings required for the issuance or continuance of a permit or license by the Commission.

70 Paperwork Reduction Act 71 This regulatory guide discusses information collection requirements covered by 10 CFR Part 20 and 72 10 CFR Part 50 that the Office of Management and Budget (OMB) approved under OMB control numbers 73 3150-0014 and 3150-0011 respectively. The NRC may neither conduct nor sponsor, and a person is not 74 required to respond to, an information collection request or requirement unless the requesting document 75 displays a currently valid OMB control number.

76 B. DISCUSSION 77 Reason for Revision 78 On December 4, 2007, the NRC revised the definition of the TEDE in 10 CFR Parts 20 and 50.

79 Under the revised rule, the TEDE means the sum of the effective dose equivalent for external exposures 80 (hereafter referred to as the EDEX) and the committed effective dose equivalent for internal exposures 81 (hereafter referred to as the CEDE). This revision of RG 8.34 provides updated regulatory guidance on 82 monitoring criteria and methods of calculating occupational dose based on the revised definition of the 83 TEDE. This regulatory guide also provides updated guidance on acceptable methods of:

84

• Determining the need for monitoring and demonstrating compliance, 85

• Monitoring alpha intakes and determining internal dose, 86

• Placement of dosimetry and resolving differences between passive and electronic 87 dosimeters, 88

• Assessing intakes and committed dose equivalent from wounds, and DRAFT RG 8.34, Revision 1, Page 3

89

• Additional calculational methods of determining internal doses.

90 Background

91 On December 4, 2007, the NRC revised the definition of the total effective dose equivalent (TEDE) 92 in 10 CFR Part 20, Standards for Protection against Radiation, Section 20.1003, Definitions and 10 CFR 93 50, Section 50.2, Definitions, (72 FR 68043 (Ref. 9)). The revision subsequently affected the methods of 94 monitoring and calculating occupational radiation doses and demonstrating compliance with the occupational 95 dose limits. Previously, the definition of the TEDE was the sum of the deep dose equivalent (DDE) to 96 account for external exposure and the committed effective dose equivalent (CEDE) to account for internal 97 exposure. Under the revised rule, 10 CFR Part 20, Section 20.1003, Definitions, the TEDE was redefined 98 by replacing the DDE with the EDEX.

99 Old definition: TEDE = DDE + CEDE 100 New definition: TEDE = EDEX + CEDE 101 In uniform radiation fields, the EDEX is normally determined by measuring the DDE and, therefore, 102 the revised TEDE definition has little impact on monitoring methods. However, for exposures in 103 non-uniform radiation fields, the revised TEDE definition provides greater monitoring flexibility and 104 accuracy for licensees in monitoring worker exposures. Under non-uniform conditions, the previous TEDE 105 definition tended to provide dose assessments that were excessively conservative.

106 Occupational dose limits are applicable during routine operations, planned special exposures, and 107 during emergencies. Doses received during declared nuclear emergencies (including international 108 emergencies) must be included in the determination of annual occupational dose. However, the potential for 109 exceeding a dose limit during a declared emergency should not prevent a licensee from taking necessary 110 actions to protect health and safety.

111 Occupational Dose Limits for Adults, Minors, and Embryo/Fetus 112 For adults, occupational dose limits (except for planned special exposures) are established in 10 113 CFR 20.1201(a) as follows:

114

• For protection against stochastic effects, the annual TEDE limit of is 5 rem (50 115 millisieverts (mSv)).

116

• For protection of adults against nonstochastic effects, the annual total organ dose 117 equivalent (TODE) limit is 50 rem (500 mSv).

118

• For protection of the lens of the eye, the annual lens dose equivalent (LDE) limit is 119 15 rem (150 mSv).

120

• For protection of the skin of the whole body or to the skin of any extremity, the annual 121 shallow-dose equivalent (SDE) limit of 50 rem (500 mSv).

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122 For minors, 10 CFR 20.1207, Occupational Dose Limits for Minors, establishes an annual limit 123 at 10 percent of the adult limits.

124 For protection of an embryo/fetus of a declared pregnant woman, 10 CFR 20.1208, Dose 125 Equivalent to an Embryo/Fetus, establishes a dose equivalent limit of 0.5 rem (5 mSv).

126 Planned Special Exposures (PSEs) 127 PSEs are subject to the conditions specified in 10 CFR 20.1206, Planned Special Exposures, 128 (e.g., exceptional circumstances, specific authorizations, and informing and instructing the worker).

129 Regulatory Guide 8.35 "Planned Special Exposures provides guidance on conducting PSEs. For dose 130 accounting purposes, dose received during a PSE is in addition to and accounted for separately from the dose 131 that is limited by 10 CFR 20.1201.

132 Surveys1 133 Surveys; i.e., evaluations of the radiological conditions and potential hazards, should be conducted 134 as necessary in support of radiological monitoring and calculation of occupational dose. Instruments and 135 equipment used in performing surveys must be calibrated periodically for the type of radiation measured in 136 accordance with 10 CFR 20.1501(c).

137 When a licensee assigns or permits the use of respiratory protection equipment to limit the intake of 138 radioactive material, 10 CFR 20.1703(c)(2) requires surveys and bioassays, as necessary, to evaluate actual 139 intakes. Indications of an intake could include facial contaminations, nasal contamination, malfunctioning 140 respiratory protection equipment, loss of engineering controls creating an airborne radioactivity area, and 141 work in unknown or unplanned airborne radioactivity areas.

142 During operations, licensees should perform airborne radioactivity surveys as required in 143 10 CFR 20.1502 to characterize the radiological hazards that may be present and, as appropriate, use 144 engineering and respiratory protection equipment to reduce intakes. When it is not practical to use process or 145 engineering controls to reduce the concentrations of airborne radioactivity to values below those that define 146 an airborne radioactivity area, licensees are required under 10 CFR 20.1702(a), consistent with maintaining 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.

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147 the TEDE ALARA, to increase monitoring (e.g., perform air sampling and Derived Air Concentrations 148 (DAC)-hour tracking and bioassay) and to limit intakes by the use of access controls, limiting exposures 149 times, or use respiratory protection equipment.

150 Monitoring At Levels Sufficient to Demonstrate Compliance 151 10 CFR 20.1502 requires monitoring at levels sufficient to demonstrate compliance with the 152 occupational dose limits; therefore monitoring methods should be reasonably accurate. Radiological surveys 153 and exposure times should be used as needed to account for dose not measured by a dosimetry system (e.g.,

154 due to dosimetry system sensitivity, or dosimeter placement, or dosimeter capability (e.g., not capable of 155 measuring minor amounts of dose from neutrons or low energy photons).

156 Licensees may voluntarily issue individual monitoring devices or use calculation methodologies 157 for reasons other than for required personnel monitoring under the requirements in 10 CFR 20.1502 (e.g., to 158 provide for worker knowledge or concern). The results of monitoring when voluntarily provided, but not 159 required by 10 CFR 20.1502, are not subject to the dose recording or reporting requirements in 10 CFR 160 Part 20, Subpart L, Records or Subpart M, Reporting. However, licensees may voluntarily provide these 161 reports to the exposed individual(s) and to the NRC.

162 Use of Effective DACs 163 10 CFR 20.1204(e) provides for methods of determining internal exposure when the identity and 164 concentration in a mixture of radionuclides are present. The identity and concentration of radionuclides may 165 be determined by surveys requiring the specific radionuclides and their relative mix. Once the relative mix is 166 known, licensees may make use of this knowledge, and apply scaling factors applicable to the mixture for use 167 in calculating DACs and tracking DAC-hours as specified in 20.1204(e). This is commonly referred as 168 effective DACs, and is applicable for beta/gamma activity, alpha activity, and hard-to-detect radionuclides.

169 The use of effective DAC values may be needed in operational radiological protection programs to 170 establish airborne radioactivity postings, alarm set points for continuous air monitors, determining the need 171 for respiratory protection, estimating internal dose, or determining when bioassay may be needed.

172 Alpha Monitoring at Nuclear Power Plants 173 For reactor facilities that have experienced significant fuel defects, alpha contamination may be a 174 radiological hazard requiring specific evaluation. Alpha contamination (when present) requires specific 175 evaluation because the alpha DACs are generally orders of magnitude more restrictive than DACs for other 176 beta-emitting and gamma-emitting isotopes.

177 Each facility should characterize and periodically update its alpha source term, based on historical 178 and current survey data and alpha spectroscopy measurements. Alpha source term characterization should not 179 be based solely on the samples of dry activated waste collected for 10 CFR 61 waste classification purposes.

180 Loose contamination surveys may not be sufficient to identify fixed alpha contamination that may be present 181 and a hazard during abrasive work (e.g., grinding, cutting or welding). A site-specific characterization should 182 determine the extent of the alpha hazard within specific areas of the plant (such as contained within localized 183 areas within the primary reactor coolant boundary or having spread to generally contaminated areas).

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184 The extent of the radiological characterization that is needed depends on the relative significance of 185 the alpha source term compared to other radiological contaminants. A site-specific alpha source term may be 186 used to identify radionuclides and determined their relative concentrations in a mixture, such as to comply 187 with the requirements of 10 CFR 20.1204(f). Once the relative concentrations are known, an effective 188 Derived Air Concentration (DAC) may be determined and used in radiological protection and dose 189 assessment.

190 Note: Methods and criteria that are acceptable for identifying and controlling alpha hazards are 191 described in the EPRI guidelines, EPRI Alpha Monitoring and Control Guidelines for Operating Nuclear 192 Power Stations, Revision 2, August, 2013 (ML14083A535) (Ref. 11).

193 The principal transuranic nuclides producing alpha radiological hazards include the isotopes of 194 curium, plutonium, and americium. For historical fuel failures (e.g., ten years since significant fuel failure),

195 the shorter-lived curium-242 has largely decayed leaving the longer-lived alpha radionuclides with more 196 restrictive DACS and ALIs as the most prevalent hazard. However, more recent fuel failures are likely to 197 identify curium-242 as the most abundant alpha emitting nuclide, which has less restrictive DAC and ALI 198 values. Therefore, the use of effective DAC values must account for the time dependent mix of alpha 199 radionuclides.

200 The extent of radiological protection measures against alpha radionuclides may be determined 201 based upon:

202

• Knowledge of the specific alpha radionuclide mix; 203

• Conservatively assuming the most restrictive radionuclide in the mixture; or 204

• Determining site specific, effective-DAC alpha values.

205 Discrete Radioactive Particle Monitoring and SDE 206 A discrete radioactive particle (DRP) is a radioactive particle that is a small, usually microscopic, 207 highly radioactive beta or beta-gamma emitting particles having relatively high specific activity. DRPs are 208 primarily an external exposure hazard to the skin, as measured by the SDE.

209 In 2002, the NRC amended its regulations related to the shallow dose equivalent/skin dose limit in 210 10 CFR Part 20 (67 FR 16298, (Ref. 12) (see also Regulatory Issue Summary 2002-10, Revision of the Skin 211 Dose Limit in 10 CFR Part 20, (Ref. 23). The amended regulations changed the definition and method of 212 calculating shallow-dose equivalents (SDE) by specifying that the assigned SDE must be the dose averaged 213 over the contiguous 10 cm2 of skin receiving the highest exposure, rather than 1 cm2 as previously 214 recommended by the NCRP (NCRP Report No. 106, Limit for Exposure to Hot Particles on the Skin (1980).

215 Harmonization with International Standards 216 The NRC has a goal of harmonizing its guidance with international standards, to the extent 217 practical. The International Commission on Radiological Protection (ICRP) and the International Atomic 218 Energy Agency (IAEA) have issued a significant number of standards, guidance and technical documents, 219 and recommendations addressing good practices in most aspects of radiation protection. The NRC DRAFT RG 8.34, Revision 1, Page 7

220 encourages licensees to consult these international documents noted throughout this guide and implement the 221 good practices, where applicable that are consistent with NRC regulations. These documents are:

222

• ICRP Publication 26, Recommendations of the International Commission on 223 Radiological Protection, (Ref. 14),

224

• ICRP Publication 30, (7-volume set including supplements), Limits for Intakes of 225 Radionuclides by Workers, (Ref. 15),

226 Documents Discussed in Staff Regulatory Guidance 227 Although this regulatory guide utilizes information, in part, from one or more reports developed by 228 external organizations and other third party guidance documents, the regulatory guide does not endorse these 229 references other than as specified in this regulatory guide. These reports and third party guidance documents 230 may contain references to other reports or third party guidance documents (secondary references). If a 231 secondary reference has itself been incorporated by reference into NRC regulations as a requirement, then 232 licensees and applicants must comply with that requirement in the regulation.

233 If the secondary reference has been endorsed in a regulatory guide as an acceptable approach for 234 meeting an NRC requirement, then the reference constitutes a method acceptable to the NRC staff for meeting 235 that regulatory requirement as described in the specific regulatory guide. If the secondary reference has 236 neither been incorporated by reference into NRC regulations nor endorsed in a regulatory guide, then the 237 secondary reference is neither a legally-binding requirement nor a generic NRC approval as an acceptable 238 approach for meeting an NRC requirement. However, licensees and applicants may consider and use the 239 information in the secondary reference, if appropriately justified and consistent with current regulatory 240 practice, consistent with applicable NRC requirements such as 10 CFR Part 20.

241 C. STAFF REGULATORY GUIDANCE 242 1. Monitoring Criteria 243 10 CFR 20.1502, Conditions Requiring Individual Monitoring of External and Internal 244 Occupational Dose, requires individual monitoring of external and internal occupational dose under the 245 radiological conditions specified below. Monitoring of external radiation exposure (i.e., the EDEX) is 246 required by 10 CFR 20.1502(a) for any individual entering a high or very high radiation area from licensed DRAFT RG 8.34, Revision 1, Page 8

247 and unlicensed2 sources under the control of the licensee. Monitoring is also required for any individual if the 248 external occupational dose is likely to exceed:

249

• For adults, 10 percent of the occupational dose limits in 10 CFR 20.1201(a).

250

• For minors in one year, a deep-dose equivalent of 0.1 rem (1 mSv), a lens dose equivalent 251 of 0.15 rem (1.5 mSv), and a shallow-dose equivalent to the skin of the whole body or to 252 the skin of the extremities of 0.5 rem (5 mSv).

253

• For declared pregnant women during the entire pregnancy, a deep-dose equivalent of 0.1 254 rem (1 mSv).

255 Monitoring the intake of radioactive material and assessing the CEDE is required by 256 10 CFR 20.1502(b) if the intake is likely to exceed:

257

• For adults, 10 percent of the applicable annual limit on intake (ALI) 258

• For minors in one year, 0.1 rem (1 mSv).

259

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

260 2. Occupational Dose 261 The definition of occupational dose, in 10 CFR 20.1003, Definitions, includes dose received 262 during the course of employment in which assigned duties involve exposure to radiation or radioactive 263 material from licensed and unlicensed sources of radiation, whether in the possession of the licensee or other 264 person. The definition of occupational dose was changed in 1995 (60 FR 36038) (Ref. 19) such that 265 occupational dose applies to workers whose assigned duties involve exposure to radiation, irrespective of 2 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; 10 CFR 40.13, Unimportant Quantities of Source Material), naturally occurring radioactive materials that are not covered by the Atomic Energy Act, radioactive materials or nuclear facilities operated by another Federal entity such as the U.S. Department of Defense or the U.S. Department of Energy; as well as machines that produce radiation, such as x-ray radiography machines and x-ray machines used by security staff.

DRAFT RG 8.34, Revision 1, Page 9

266 their location inside or outside a restricted area. Note: A member of the public does not become an 267 occupationally exposed individual as a result of just entering a restricted area.

268 Individuals who receive occupational exposure and are likely to receive more than 100 mrem must 269 be instructed in accordance with 10 CFR 19.12. See Regulatory Guide 8.29, Instruction Concerning Risks 270 from Occupational Radiation Exposure for further information.

271 3. Prospective Assessments of the Need for Occupational Dose Monitoring 272 Licensees must identify those individuals receiving occupational dose, either individually or 273 as a group or category of individuals. Individuals pre-designated by the licensee as receiving occupational 274 dose are subject to the occupational dose limits; otherwise, individuals must be considered as members of the 275 public subject to public dose limits in 10 CFR 20.1301, Dose limits for individual members of the public.

276 Once occupationally exposed individuals are identified, licensees should perform a 277 prospective assessment to determine if those individuals are likely to exceed the minimum exposure levels 278 specified in 10 CFR 20.1502 (i.e., determine the need for monitoring of the occupational dose). As discussed 279 in 60 FR 36039 (1995) (Ref. 22), the term likely to receive includes normal situations as well as abnormal 280 situations involving exposure to radiation which can reasonably be expected to occur during the life of the 281 facility. Reactor licensees should consider normal operations and anticipated operational occurrences (e.g.,

282 unplanned onsite events, such as sudden increases in external radiation levels, or localized high airborne 283 radioactivity areas) but would not need to consider design basis accidents.

284 The prospective assessment determines the type of monitoring required (e.g., external dose 285 or internal dose monitoring). In performing a prospective assessment, an evaluation should be performed 286 based on planned work activities and likely exposure conditions. Prospective assessments should be revised 287 when there are substantial changes to the radiological conditions of personnel exposure (e.g., changes in work 288 activities, airborne concentrations, beta energy spectrums, or use of new or different types or energies of 289 radiation producing equipment.)

290 The requirements for monitoring in 10 CFR 20.1502 refers to exposures that may occur at 291 each licensee individually. Doses that have already been received under another licensee, or may be 292 received in the future from employment by another licensee or unlicensed entity, are excluded from 293 consideration in a licensees determination of the need to monitor an individual. The need for monitoring 294 should be based on the anticipated exposure to licensed or unlicensed sources under the control of a single 295 licensee.

296 4. Determination of External Doses 297 a. Determination of the TEDE 298 Under 10 CFR 20.1202, if a licensee is required to monitor both external dose and internal 299 dose, the licensee must demonstrate compliance with the dose limits by summing external and internal doses 300 (i.e., TEDE = EDEX + CEDE). However, if the licensee is only required to monitor external doses under 301 10 CFR 20.1502(a), or only internal doses under 10 CFR 20.1502(b), then summation is not required to 302 demonstrate compliance with the occupational dose limits. For example, if the internal dose is not DRAFT RG 8.34, Revision 1, Page 10

303 monitored, the CEDE can be assumed equal to zero, and the TEDE is equal to the EDEX. Similarly, if the 304 external dose is not monitored, the EDEX can be assumed equal to zero, and the TEDE is equal to the CEDE.

305 b. Determination of the EDEX 306 The EDEX is determined using one or more combinations of the following methods 307 in accordance with 10 CFR 20.1201(c). These methods are described in RG 8.40 as follows:

308

• Measuring the DDE at the highest exposed part of the whole body with an external 309 personal monitoring device, as required by 10 CFR 20.1201(c), when an NRC method 310 for determining EDEX is not used.

311

• Measuring external exposure with one or more external personal monitoring devices and 312 determining EDEX using an NRC approved method such as those provided in 313 Regulatory Guide 8.40, or as specifically approved by the NRC.

314

• Calculating the EDEX based on survey data obtained under 10 CFR 20.1501 or other 315 radiological data, such as known source activity, dose rates, and exposure times using 316 scientifically sound technical methods. This may be required under unique exposure 317 situations (e.g., partial body exposed to radiation streaming of narrow beam geometries) 318 or when the individual monitoring device was not in the region of the highest whole body 319 exposure (per 10 CFR 20.1201(c)), or the results of the individual monitoring are not 320 available (i.e., damaged or lost device).

321 Note: Within the same monitoring period, a licensee may use a combination of methods 322 above; e.g., a licensee may routinely determine EDEX for the majority of a monitoring period using 323 method (1) above, and then use the methods (2) and/or (3) for special exposure situations at other 324 times. The results of the different dosimetry methods must be combined to determine the EDEX for 325 the entire monitoring period.

326 c. Determination of the Deep-Dose Equivalent (DDE) 327 The DDE is typically measured by the use of an individual monitoring device(s) and is 328 determined at a tissue depth of 1 centimeter (cm) (1,000 mg/cm2). The DDE can also be calculated if 329 the appropriate parameters (i.e., radiation source strength, exposure geometry, full or partial 330 shielding) are known.

331 A single DDE located at the highest exposed part of the whole body is a conservative (and for 332 uniform exposures, a reasonably accurate) estimate of the EDEX from external sources. However, 333 there are several other NRC approved methods for determining EDEX provided in RG 8.40 that use 334 external monitoring devices measuring DDE at specific locations on the whole body. See the RG 335 8.40 for the use and limitations of each method.

336 In many exposure situations, the passive dosimeter used for the single DDE measurement 337 may be supplemented with an active dosimeter (e.g., electronic dosimeter) for work control or dose 338 accounting purposes (i.e., an active dosimeter provides real time indication of the accrued dose and DRAFT RG 8.34, Revision 1, Page 11

339 possibly the dose rate). Due to the differences in dosimeter design and detection technology (and the 340 relative measurement errors associated with each) there can be differences in reading of these two 341 dosimeters for the same exposure, even if the dosimeters are co-located on the monitored individual.

342 Within a reasonable pre-established accuracy criteria (depending of the dosimeter designs), the small 343 differences can be disregarded and either dosimeter value used as the measured dose (i.e., the 344 readings are considered the same value). However, if dosimeter readings are outside the established 345 accuracy criteria, and unresolved, then the highest reading must be recorded as the dose received 346 during the exposure period per the requirement in 20.1201(c).

347 d. Determining the LDE 348 The LDE is defined at a tissue depth of 0.3 cm (300 mg/cm2). If the LDE is being monitored 349 with a dosimeter, then that dosimeter should be calibrated to measure the dose at a tissue depth of 350 7 mg/cm2. Alternately, the LDE may be conservatively determined based on SDE measurements. In 351 many exposure situations, shield glasses can be worn to prevent exposures to the lens of the eye from 352 low energy (or low penetrating) radiations, eliminating the need for monitoring the LDE.

353 e. Determination of the SDE 354 The SDE is defined as the external exposure of the skin of the whole body or extremities, 355 which can result from external skin contamination such as from radioactive solids, discrete 356 radioactive particles (hot particles), or liquids on the surface of the skin or on protective clothing.

357 The SDE is defined only for external exposure at a tissue depth of 0.007 cm (7 mg/cm2), and is the 358 dose averaged over the contiguous 10 cm2 of skin receiving the highest exposure. If the SDE is being 359 measured with a dosimeter, then that dosimeter should be calibrated to measure the dose at a tissue 360 depth of 7 mg/cm2. The latest version of NUREG/CR-6918, VARSKIN: A Computer Code for 361 Assessing Skin Dose from Skin Contamination can be used to assess SDE.

362 The SDE for exposure to submersion class radionuclides containing low energy betas are not 363 readily measureable by direct survey techniques or dosimetry methods, and hence may need to be 364 calculated based on air sample analyses and DAC-hr tracking. This submersion exposure 365 information may be needed for informing workers of radiological exposure conditions (e.g., SDE 366 rates used for pre-job briefings), and also to account for the SDE that may not be adequately 367 measured by dosimeters because of the dosimeter lack of response to low energy beta spectrums.

368 5. Determination of Intakes 369 For those licensees determining internal dose per 10 CFR 20.1204, a determination must be 370 made of the intake that can occur through inhalation, ingestion, absorption through the skin, or through 371 wounds. The amount of the intake may be assessed from suitable and timely measurements of airborne 372 radionuclides or may be based on bioassay measurements.

373 The assessment of intake should include the readily-detected radionuclides as well as the 374 hard-to-detect radionuclides (not directly measured). The activity of hard-to-detect radionuclides may be 375 based on scaling factors correlated to the amount of readily-detected radionuclides. See Regulatory Guide DRAFT RG 8.34, Revision 1, Page 12

376 8.25, Air Sampling in the Workplace and Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, 377 and Assumptions for a Bioassay Program for further guidance on determining uptakes and intakes.

378 Unless respiratory protection is used, the concentration of radionuclides in the intake (i.e., the 379 breathing zone concentration) is assumed to be equal to the ambient concentration. Therefore, selecting the 380 air sample location should consider engineered features such as containment, airflow, and filtration, to ensure 381 that the air sample is representative of the air breathed.

382 If respiratory protection is used to limit the intake of radioactive materials, 10 CFR 383 20.1703(c)(4)(i) requires internal monitoring be implemented as part of the respiratory protection program.

384 When respiratory protection is provided, the intake is adjusted by dividing the ambient air concentration by 385 the appropriate Assigned Protection Factor (APF) listed in 10 CFR 20, Appendix A. If the ambient air 386 concentration is determined by performing breathing zone air sampling inside the respiratory protective 387 device (such as with a lapel air sampler inside a loose fitting supplied air hood or suit), then no APF 388 adjustment is made to the ambient air concentration as measured by the breathing zone air sample.

389 a. Determining the Intake based on Air Sampling 390 Intakes (I) based on air sampling results can be assessed by multiplying the airborne 391 concentration (C) by the breathing rate and the exposure time.

392 I = C Air sample (µCi/ml)

• breathing rate (ml/minutes)

• exposure time (minutes);

393 where the breathing rate of "Reference Man" under light working 394 conditions is 2E+4 ml/minute.

395 The intake of radionuclides can also be estimated by DAC-Hour tracking in which the 396 ambient airborne concentration (expressed as a fraction of the DAC) is multiplied by exposure time 397 (expressed in hours).

398 b. Determining the Intake based on Bioassay Measurements 399 Another method of assessing the intake from inhalation, ingestion or skin absorption is based 400 on bioassay measurements of the uptake. The can be determined based on measurements of uptakes, 401 an evaluation of the mode of intake (inhalation, ingestion or wounds), and follow-up bioassay 402 measurements to determine the retention/elimination rates. Time and motion conditions may support 403 assessments of intake as well. Guidance on methods of estimating intake based on bioassay 404 measurements of update is provided in NUREG/CR-4884, Interpretation of Bioassay 405 Measurements, (Ref. 24).

406 The intake(s) from wounds is generally assessed based on bioassay measurements using a 407 combination of whole body in vivo bioassay and hand-held instrumentation. The bioassay 408 measurements should determine the location of the injected source, such that CDE dose calculations 409 may be made to the highest exposed 10 cm2 area of the skin at a depth of 0.007 cm (see section 7.d 410 below).

DRAFT RG 8.34, Revision 1, Page 13

411 Note: The amount of the intake may be assessed using newer, updated biokinetic models 412 (e.g., ICRP Publications 60, 1990 Recommendations of the International Commission on 413 Radiological Protection, and ICRP Publication 103, The 2007 Recommendations of the 414 International Commission on Radiological Protection). However, the CEDE must be calculated 415 using the existing 10 CFR 20.1003 organ weighting factors (unless the use of other weighting factors 416 have been specifically approved by the NRC). In other words, the use of more recent tissue or organ 417 dose weighting factors is not acceptable (since the regulations in 10 CFR Part 20 list the specific 418 organ dose weighting factors that must be used).

419 c. Determining Alpha Intakes 420 Alpha intakes may be assessed based on radiological surveys and on a site-specific alpha 421 source term. After the relative concentrations of alpha emitting isotopes are determined (e.g., by 422 alpha spectroscopy), scaling factors for alpha to beta/gamma activity may be used to determine the 423 alpha activity. Scaling factors based on surface area contamination or air samples should be 424 representative of work area at the time of exposure.

425 Internal doses may be determined based on whole body count data and scaling factors when 426 nominal alpha doses occur, such as less than 500 mrem CEDE. However, if an alpha intake 427 exceeding a nominal level is considered likely, excreta sampling or lung counting may be needed to 428 determine intakes and assign dose. When excreta sampling is to be initiated, sampling should begin 429 as soon as possible following detection of the exposure, and continue for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or until at 430 least one sample is collected (following the first void for urine). ANSI N13.39 (2011), Design of 431 Internal Dosimetry Programs provides additional guidance on excreta sampling.

432 6. Determination of Internal Dose 433 a. Calculation of the Committed Effective Dose Equivalent (CEDE) 434 The dose quantity for protection against stochastic effects of internal dose is the CEDE; i.e., a 435 50-year committed effective dose equivalent from intakes occurring during the monitoring period. There are 436 three fundamental methods described below for calculating the CEDE:

DRAFT RG 8.34, Revision 1, Page 14

437

• Using dose coefficients3 from the U.S. Environmental Protection Agencys Federal 438 Guidance Report No. 11 (FGR-11) (Ref. 25).

439

• Using ALI methods.

440

• Using DAC-hour methods.

441 Details and examples on calculating the CEDE are described in Appendix A.

442 Note: When performing CEDE calculations using the ALI and DAC-hour methods, the ALI 443 and DAC values provided in Appendix B to 10 CFR Part 20 must be used, unless the licensee has 444 obtained prior NRC approval in accordance with 10 CFR 20.1204(c)(2) to adjust the ALI or DAC 445 values.

446 b. Calculation of the Committed Dose Equivalent (CDE) 447 The CDE is the 50-year committed dose equivalent from intake of radioactive material.

448 Methods and examples of calculating the CDE are described in Appendix A. The special case of 449 calculating the CDE from wound intakes is discussed in Section C.7.d below.

450 c. Calculation of the Total Organ Dose Equivalent (TODE) 451 The dose limit for protection against the nonstochastic effects is expressed in terms 452 of the TODE; i.e., the sum of the DDE and the CDE.

453 TODE = DDE + CDE 454 The TODE is determined by adding the DDE (measured at the highest exposed part of the 455 whole body) to the CDE.

456 If only internal monitoring is being performed, the TODE is equal to the CDE to the highest 457 exposed organ (since the DDE was not monitored and is assumed equal to zero). Further details on 458 acceptable methods of calculating the CDE are described in Appendix A.

3 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 since the terminology is not incorporated into the regulations).

DRAFT RG 8.34, Revision 1, Page 15

459 If both internal and external monitoring are being performed, the licensee must demonstrate 460 that both the 5 rem TEDE and the 50 rem TODE limits are met. One method of demonstrating 461 compliance with the TODE limit is by summing the DDE and the CDE to the highest exposed organ.

462 Another acceptable method of demonstrating that the TODE limit is met is by maintaining the DDE 463 to less than 5 rem, and the CEDE to less than 1 rem4, then the TODE cannot exceed the 50 rem TODE 464 limit. In this case, the CDE does not need to be determined since compliance was demonstrated by 465 calculation. If the CEDE does exceed 1 rem, the CDE must be determined in order to demonstrate 466 compliance with the dose limits.

467 d. Doses from Intakes through Wounds 468 In accordance with 10 CFR 20.1202(d), the licensee shall evaluate and, to the extent 469 practical, account for intakes through wounds.

470 10 CFR 20.1201 also specifies two annual dose limits:

471

• TODE limits (Section 20.1201(a)(1)(ii)) - the sum of the DDE and the CDE to any 472 individual organ or tissue other than the lens of the eye) being equal to 50 rem (0.5 Sv)),

473 and 474

• SDE limits (Section 20.1201(a)(2)(ii)) - the SDE of 50 rem (0.5 Sv) to the skin of the 475 whole body or skin of any extremity.

476 However, because the SDE is defined only for external exposure, the SDE limit is not 477 applicable (to dose from wound intakes). Therefore, the TODE dose limit becomes the only 478 applicable limit; i.e., a CDE limit of 50 rem to any individual organ (e.g., skin). Note: In most skin 479 exposure situations, the dose is from external exposure (and therefore the dose to the skin organ is 480 commonly equal to the SDE). However, when the dose to the skin (organ dose) is from a wound, the 481 CDE (organ) dose limit applies (not the SDE).

482 In making the TODE dose calculation (to the skin organ) under 20.1201(a)(1)(ii), the DDE 483 component is zero, since for intakes by wounds, the DDE is zero (since DDE is an external 484 whole-body exposure). As a result, the calculated dose is only the CDE to the skin calculated to the 4

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.

DRAFT RG 8.34, Revision 1, Page 16

485 highest exposed, contiguous 10 cm2 area at a depth of 0.007 cm (in a manner similar to SDE 486 calculations).

487 In summary, the CDE to the skin is the appropriate quantity to be calculated (50-year 488 integrated dose (until the source is removed), at a depth of 0.007 centimeters below the surface of the 489 skin, and averaged over the highest exposed 10 cm2 of the basal layer of the skin. In order to do this 490 calculation, the location (depth) of the source must be determined as an input parameter, and the most 491 recent version of Varskin computer code may be used in performing calculations.

492 For wound intakes with systemic uptakes, an evaluation must be performed of the CEDE and 493 TEDE. Additional information on assessing intakes through wounds is available ICRP-54 (Ref. 26),

494 ICRP-78 (Ref. 27), NCRP-87 (Ref. 28), and technical articles by Toohey (Ref. 29) and Ishique (Ref.

495 30).

496 Note: With respect to tissue dose, there is no regulatory limit for small volume, localized 497 tissue dose. However, licensees should estimate the committed dose to underlying tissues (e.g., 1 498 cm3 of flesh) at the wound site for purposes of determining the potential for tissue function 499 impairment and whether medical intervention is warranted (e.g., surgical removal). The guidance in 500 NCRP Report No. 156, Development of a Biokinetic Model for Radionuclide-Contaminated 501 Wounds and Procedures for Their Assessment, Dosimetry, and Treatment is acceptable for this 502 evaluation (Ref. 31).

503 e. Calculating the CDE and CEDE for Inhalation, Submersion and Absorption 504 A number of methods are acceptable for calculating the CDE and CEDE from the intake of 505 radioactive materials. Some of these methods are described below. However, calculations of the 506 CEDE must be based on the 10 CFR Part 20 organ weighting factors and specified tissues. The more 507 recent ICRP Publication 68 dose coefficients cannot be used, (unless their use has been specifically 508 approved by the NRC). This is because the ICRP 68 and ICRP 103 tissues and weighting factors are 509 different from those in 10 CFR Part 20.

510 7. Use of Individual or Material-Specific Information 511 The regulation at 10 CFR 20.1204(c) states that when specific information on the physical 512 and biochemical properties of the radionuclides taken into the body or the behavior of the material in 513 an individual is known, the licensee may...use that information to calculate the committed effective 514 dose equivalent.... Prior NRC approval is not required, but detailed records must be kept to 515 demonstrate the acceptability of the dose assessment.

516 The characteristics most amenable to such individual or site-specific consideration are the 517 activity median aerodynamic diameter (AMAD) of the inhaled aerosol and the solubility of the 518 material in the lungs and in the GI tract. The use of specific information on the physical and 519 biochemical properties to calculate the CEDE requires the licensee to do considerably more work and 520 to have greater technical expertise than the other methods, and therefore, this method may not be 521 useful for small, infrequent intakes. Conversely, the use of specific information of the physical and 522 biochemical properties of radionuclides taken into the body may be appropriate in the case of DRAFT RG 8.34, Revision 1, Page 17

523 accidental large exposures if more accurate information would lead to a better estimate of the actual 524 dose.

525 8. Uranium Intake Limitation 526 In accordance with 10 CFR 20.1201(e), in addition to the annual dose limits, the licensee 527 shall limit the soluble uranium intake by an individual to 10 mg in a week, in consideration of chemical 528 toxicity. Regulatory Guide 8.11, Applications of Bioassay for Uranium describes methods acceptable for 529 the design of bioassay programs for protection against intake of uranium, conditions under which bioassay is 530 necessary, minimum quantifiable values for direct and indirect bioassay measurements, protection 531 guidelines, and objectives.

532 D. IMPLEMENTATION 533 The purpose of this section is to provide information to applicants and licensees regarding the 534 NRCs plans for using this regulatory guide.

535 Methods or solutions that differ from those described in this regulatory guide may be deemed 536 acceptable if they provide sufficient basis and information for the NRC staff to verify that the 537 proposed alternative complies with the appropriate NRC regulations. Current licensees may continue 538 to use guidance the NRC found acceptable for complying with the identified regulations as long as 539 their current licensing basis remains unchanged.

540 541 542 543 DRAFT RG 8.34, Revision 1, Page 18

544 REFERENCES 545 1. 10 CFR Part 20, Standards for Protection against Radiation, U.S. Nuclear Regulatory Commission, 546 Washington, DC 20555.

547 2. 10 CFR 50, Section 50.2, Definitions, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

548 3. 10 CFR Part 19, Notices, Instructions, and Reports to Workers: Inspection and investigations, 549 U.S. Nuclear Regulatory Commission, Washington, DC 20555.

550 4. Regulatory Guide 8.9, Revision 1, Interpretation of Bioassay Measurements U.S. Nuclear Regulatory 551 Commission, Washington, DC 20555.

552 5. Regulatory Guide 8.25, Revision 1, Air Sampling in the Workplace U.S. Nuclear Regulatory 553 Commission, Washington, DC 20555.

554 6. Regulatory Guide 8.35, Revision 1, "Planned Special Exposures U.S. Nuclear Regulatory Commission, 555 Washington, DC 20555.

556 7. Regulatory Guide 8.36, Radiation Dose to the Embryo/Fetus U.S. Nuclear Regulatory Commission, 557 Washington, DC 20555.

558 8. Regulatory Guide 8.40, Methods for Measuring Effective Dose Equivalent from External Exposure, 559 U.S. Nuclear Regulatory Commission, Washington, DC 20555.

560 9. Federal Register Notice 72 FR 68043 Occupational Dose Records, Labeling Containers, and the Total 561 Effective Dose Equivalent, U.S. Nuclear Regulatory Commission, Washington, DC, December 2007.5 562 10. Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure 563 Data U.S. Nuclear Regulatory Commission, Washington, DC 20555.

564 11. EPRI Alpha Monitoring and Control Guidelines for Operating Nuclear Power Stations, Revision 2, 565 August, 2013 (ML14083A535) (Ref.11) 9 Printed copies of Federal Register notices are available for a fee from the U.S. Government Printing Office, 732 N Capitol Street, NM 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/.

DRAFT RG 8.34, Revision 1, Page 19

566 12. Federal Register Volume 67, Revision of the Skin Dose Limit (pp.16298-16301) U.S. Nuclear 567 Regulatory Commission, Washington, DC, April 2002.

568 13. International Commission on Radiological Protection (ICRP) Publication 2, Permissible Dose for 569 Internal Radiation, International Commission on Radiological Protection, Pergamon Press, Oxford, UK, 570 1959 571 14. ICRP Publication 26, Recommendations of the International Commission on Radiological Protection, 572 Protection, Pergamon Press, Oxford, UK, 1977.

573 15. ICRP Publication 30, (7-volume set including supplements), Limits for Intakes of Radionuclides by 574 Workers, International Commission on Radiological Protection, Pergamon Press, Oxford, UK, 1982.

575 16. ICRP Publications 60, 1990 Recommendations of the International Commission on Radiological 576 Protection, International Commission on Radiological Protection, Pergamon Press, Oxford, UK, 1990.

577 17. ICRP Publication 68, Dose Coefficients for Intakes of Radionuclides for Workers, International 578 Commission On Radiological Protection, Pergamon Press, Oxford, UK, 1994.

579 18. ICRP Publication 103, The 2007 Recommendations of the International Commission on Radiological 580 Protection, International Commission on Radiological Protection, Pergamon Press, Oxford, UK, 2007.

581 19. IAEA Safety Standard Series No. RS-G-1.1, Occupational Radiation Protection, , International 582 Atomic Safety Agency, Vienna, Austria, 1999.

583 20. IAEA Safety Standard Series No. RS-G-1.2, Assessment of Occupational Exposure Due To Intakes of 584 Radionuclides, , International Atomic Safety Agency, Vienna, Austria, 1999, and 585 21. IAEA Safety Standard Series No. RS-G-1.3, Assessment of Occupational Exposure Due To External 586 Sources of Radiation, , International Atomic Safety Agency, Vienna, Austria, 1999.

587 22. Federal Register Notice 60 FR 36038 Radiation Protection Requirements: Amended Definitions and 588 Criteria, U.S. Nuclear Regulatory Commission, Washington, DC, July 1995.

589 23. Regulatory Issue Summary 2002-10, Revision of the Skin Dose Limit in 10 CFR Part 20, U.S. Nuclear 590 Regulatory Commission, Washington, DC (ADAMS ML021860332).

591 24. NUREG/CR-4884, Interpretation of Bioassay Measurements, U.S. Nuclear Regulatory Commission, 592 Washington, DC, June 1990 (ADAMS ML11285A018).

593 25. U.S. Environmental Protection Agencys Federal Guidance Report No. 11 (FGR-11) (Ref. 25).

594 26. ICRP Publication 54, Individual Monitoring for Intakes of Radionuclides by Workers, Chapters 4.2 &

595 4.3, 1989.

596 27. ICRP Publication 78, Individual Monitoring for Internal Exposure of Workers Chapter 4.2, 1997.

DRAFT RG 8.34, Revision 1, Page 20

597 28. NCRP Report 87, NCRP Report No. 87, Use of Bioassay Procedures for Assessment of Internal 598 Radionuclide Deposition, Chapters 5.3.1, 5.3.2, 5.4.6 599 29. Toohey , R.E., et al, "Dose Coefficients for Intakes of Radionuclides Via Contaminated Wounds, Health 600 Phys. 2011 May;100(5):508-14; also available from the Oak Ridge Institute for Science and Education at 601 URL: http://orise.orau.gov/reacts/resources/retention-intake-publication.aspx 602 30. Ishique, N., "Implementation of the NCRP Wound Model for Interpretation of Bioassay Data for Intake 603 of Radionuclides Through Contaminated Wounds", Radiation Research. 2009 May; 50(3): 267-276 604 31. NCRP Report No. 156, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds 605 and Procedures for Their Assessment, Dosimetry, and Treatment National Council on Radiation 606 Protection and Measurements, Library of Congress, Washington, DC.

607 32. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory 608 Commission, Washington, DC 20555.

609 33. ANSI N13.39 (2011), Design of Internal Dosimetry Programs 610 34. Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay 611 Program, U.S. Nuclear Regulatory Commission, Washington, DC.

612 35. Regulatory Guide 8.11, Applications of Bioassay for Uranium 613 36. Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation Exposure 614 37. Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure 615 Data, U.S. Nuclear Regulatory Commission, Washington, DC.

616 38. K. F. Eckerman, A. B. Wolbarst, and A. C. B. Richardson, Limiting Values of Radionuclide Intake and 617 Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal 618 Guidance Report No. 11 (EPA 520/1-8-020), U.S. Environmental Protection Agency, Washington, DC, 619 1988 620 39. Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay 621 Program, Chapter 3, page 8.9-4 622 40. Information Notice No. 97-36: (June 27,1997), Unplanned Intakes by Worker of Transuranic Airborne 623 Radioactive Materials and External Exposure due to Inadequate Control of Work 624 41. SECY-98-245 - Rulemaking Plan - Protection Against Discrete Radioactive Particle" 625 42. NUREG/CR-6918, VARSKIN 4: A Computer Code for Assessing Skin Dose from Skin 626 Contamination 627 43. NCRP Report No. 106, Limit for Exposure to Hot Particles on the Skin (1980) 628 DRAFT RG 8.34, Revision 1, Page 21

629 Appendix A 630 631 1. Calculations of the CDE and the CEDE for Any Radionuclide, based on Bioassay 632 Measurements using the Dose Coefficients from Federal Guidance Report No. 11 633 This method is based on using tabulated dose coefficients to calculate the dose. The FGR-11 634 provides tables of dose coefficients (DCs) (FGR-11 uses the terminology dose conversion factors) 635 for intakes by inhalation and by ingestion (see excerpt below for inhalation of Co-60). FGR-11 636 provides two types of DCs:

637 638 (1) DCs for the CDE to an organ or tissue per unit of activity (DCorgan) (e.g. the heading 639 Lung below) and 640 641 (2) DCs for the CEDE per unit of activity (DCeffective) (as shown in the far right column of 642 the tables under the heading Effective) 643 644 If site-specific information is known about the type of compound and its clearance class, the 645 appropriate clearance class can be selected. If not, the class is normally selected based on the most 646 conservative Class (in Example A, the DC for the lung is selected from clearance Class Y having a 647 value of 3.45E-7). Multiplying the DCs by the intake (I) for that radionuclide calculates the CDE and 648 CEDE for that radionuclide.

649 650 CDE (rem) = DCorgan (rem/µCi)

• I (µCi) 651 CEDE (rem) = DCeffective (rem/Ci)

• I (Ci) 652 653 Example 1: Calculations of the CDE and the CEDE for Co60, based on bioassay measurements 654 using the DCs from FGR-11.

655 An intake by inhalation was estimated by a whole body count to be 360 nCi (0.36 µCi) of 656 Co-60, Class Y aerosol. Calculate the CDE to the lung and the CEDE.

DRAFT RG 8.34, Revision 1, Page 22

657 From Table 2.1 of FGR-11 (see excerpt below), the DCs for Class Y, Co-60 radionuclide are 658 3.45E-7 Sv/Bq for the CDE and 5.91E-8 Sv/Bq for the CEDE.

659 660 Excerpt from Federal Guidance Report No. 11 661 662 The DCs in FGR-11 are tabulated in Sv/Bq and may be converted to mrem/Ci by 663 multiplying by 3.7x109.

664 665 DClung = (3.45E-7 Sv/Bq) * (3.7E+9) = 1,277 mrem/µCi 666 DCeffective = (5.91E-8 Sv/Bq) * (3.7E+9) = 219 mrem/µCi 667 668 The doses are calculated by multiplying these DCs by the intake of 0.36 µCi.

669 670 CDElung = (1,277 mrem/µCi) * (0.36 µCi) = 460 mrem 671 CEDE = (219 mrem/µCi) * (0.36 µCi) = 79 mrem 672 673 2. Calculation of the CEDE based on Bioassay Measurements using Stochastic ALIs 674 The ALI values are listed in Table 1 of Appendix B to 10 CFR Part 20. Column 1 lists the 675 values for oral ingestion, and Column 2 lists the values for inhalation. The stochastic ALI values can 676 be used in the calculation of the CEDE, based on the fraction of the allowable annual intake and the 5 677 rem (50 mSv) CEDE dose limit. When the ALI is defined by the stochastic limit, this value alone is 678 given in the table.

679 680 681 DRAFT RG 8.34, Revision 1, Page 23

682 Since the stochastic ALI corresponds to a 5 rem (50 mSv) CEDE dose limit, the CEDE may 683 be calculated based on the ratio of the intake to the stochastic ALI, multiplied by 5 rem (50 mSv).

684 685 CEDE = (I/ALI)

• 5 rem 686 687 Example 2: Calculate the CEDE based on bioassay measurements using the stochastic ALIs.

688 The intake by inhalation for a worker was estimated by bioassay to be 360 nCi (0.36 µCi) of 689 Co-60, Class Y aerosol. Calculate the CEDE.

690 691 From Appendix B above, Table 1, Column 2, the ALI for Class Y Co-60 is:

692 693 ALI (stochastic) = 30 µCi 694 CEDE = (I/ALI)

• 5 rem 695 CEDE = (0.36 µCi/30 µCi)

• 5 rem = 0.06 rem = 60 mrem 696 697 Note: Considering the precision of a 1 significant figure for the ALI values, this 60 mrem 698 value compares favorably to the calculated CEDE value of 79 mrem determined in Example 699 A above using the FGR-11 method. Either calculational method and/or result is acceptable in 700 demonstrating compliance.

701 702 3. Calculation of the CDE based on Bioassay Measurements Using Nonstochastic ALIs 703 The nonstochastic ALI values can be used in the calculation of the CDE, based on the 704 fraction of the allowable annual intake and the 50 rem (500 mSv) CDE dose limit. When the ALI is 705 defined by the nonstochastic limit, this value is listed first in the table with its corresponding organ 706 (see excerpt below), and the corresponding stochastic ALIs are given in parenthesis below (e.g.,

707 9E+1 µCi for ingestion and 2E+2 µCi for inhalation in excerpt below).

708 709 710 711 DRAFT RG 8.34, Revision 1, Page 24

712 Since the nonstochastic ALI corresponds to a 50 rem (500 mSv) CDE dose limit, the CDE 713 may be calculated based on the ratio of the intake to the nonstochastic ALI, multiplied by 50 rem (500 714 mSv).

715 CDE = (I/ALI)

• 50 rem 716 717 Note: For a mixture of radionuclides, the sum of the fractions technique as described in 10 CFR 718 20.1202(b) must be used.

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

720 721 The intake by inhalation for a worker was estimated by bioassay to be 131 nCi (0.131 µCi) of 722 I-131, Class D aerosol. Calculate the CDE to the thyroid.

723 724 From Appendix B above, Table 1, Column 2, the ALI for Class D I-131 is:

725 726 ALI (nonstochastic) = 5E+1 µCi = 50 µCi 727 CDE = (0.131 µCi/50 µCi)

• 50 rem = 0.131 rem = 131 mrem 728 729 4. Calculation of the CDE based on air sampling and nonstochastic DAC-hrs 730 For nonstochastic radionuclides, an exposure to an airborne concentration of 1 DAC results 731 for a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> exposure time results in 50 rem CDE; or 50,000 mrem/2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, or 25 mrem CDE 732 per DAC-hour.

733 CDE = [25 mrem per DAC-hr]

• number of DAC-hrs 734 where the number of DAC-hrs = (air concentration / DAC value)

• exposure time 735 Example 4: Calculation of the CDE based on air sampling and nonstochastic DAC-hrs.

736 737 Calculate the CDE to the thyroid for a 30-minute exposure based on an air sample result of 738 2.1E-7 µCi/ml (I-131).

739 740 The nonstochastic DAC for I-131 is listed in Appendix B (see excerpt below) as 2E-8 741 µCi/ml.

742 DRAFT RG 8.34, Revision 1, Page 25

743 744 745 CDE = 25 mrem/DAC-hr * [(2.1E-7 µCi/ml) / (2E-8 µCi/ml)] number of DACs * (0.5 hrs) = 131 746 mrem 747 5. Calculations of the CEDE based on air sampling and stochastic DAC-hrs 748 For stochastic radionuclides (e.g., Co-60), an exposure to an airborne concentration of 1 DAC results 749 in 5,000 mrem CEDE in 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of exposure time; or 5,000 mrem/2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, or 2.5 mrem 750 CEDE per stochastic DAC-hr.

751 CEDE = [(2.5 mrem)/DAC-hr]

• No. of DAC-hrs 752 where the number of DAC-hrs = (air concentration / DAC value)

• exposure time 753 Example 5: Calculation of the CEDE based on air sampling and stochastic DAC-hrs.

754 755 Calculate the CEDE for a 30-minute exposure based on an air sample result of 2.1E-7 µCi/ml 756 (Co-60).

757 758 From Appendix B below, the stochastic DAC for Co-60 clearance Class Y compound is 1E-8 759 µCi/ml.

760 761 762 CEDE = [2.5 mrem/DAC-hr] * [(2.1E-7 µCi/ml) / (1E-8 µCi/ml)] No. of DACs* (0.5 hrs) = 26 mrem DRAFT RG 8.34, Revision 1, Page 26

763 6. Calculation of the CEDE based on air sampling and calculated stochastic DAC-hrs 764 CEDE = [2.5 mrem/DAC-hr]

• No. of DAC-hrs 765 No. DAC-hrs = [air concentration / calculated DAC value] * [exposure time]

766 Note: Appendix B to 10 CFR Part 20 does not list the stochastic DAC values (see empty 767 circled cell below) for radionuclides with intakes limited by the nonstochastic limits.

768 However, the stochastic DAC values may be calculated based on the stochastic ALI values.

769 These stochastic ALI values are listed (in parenthesis) below the limiting nonstochastic organ 770 (see circled value of 2E+2 µCi in the table below).

771 772 Example 6: Calculation of the CEDE based on air sampling and calculated stochastic DAC-hrs.

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

774 The stochastic DAC value is first calculated by dividing the stochastic ALI by the breathing rate of 775 2.4E+9 ml/yr.

776 The calculated stochastic DAC (I-131) = (2E+2 µCi) / (2.4E+9 ml/yr) = 8E-8 µCi/ml or µCi/cc (since 777 1 ml = 1 cc) 778 CEDE = [2.5 mrem/hr/DAC-hr] * [(2.1E-7 µCi/ml) / (8E-8 µCi/ml)] DACs * (0.5 hrs) = 3.3 mrem 779 DRAFT RG 8.34, Revision 1, Page 27