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 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]

DRAFT REGULATORY GUIDE 8.34 1

(Draft was issued as DG-8031) 2 3

MONITORING CRITERIA AND METHODS TO CALCULATE 4

OCCUPATIONAL RADIATION DOSES 5

A. INTRODUCTION 6

Purpose 7

This guide provides methods acceptable to the staff of the U.S. Nuclear Regulatory Commission 8

(NRC) for monitoring the occupational radiation dose to individuals and for calculating occupational 9

radiation doses. The Regulatory Guide applies to both reactor and materials licensees under both NRC and 10 Agreement State licenses.

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

18 Also, Section 20.1201, Occupational Dose Limits for Adults, establishes radiation dose limits for 19 occupationally exposed individuals. These limits apply to the sum of the dose received from external 20 exposure and the dose from internally deposited radioactive material. Conditions that require individual 21 monitoring of external and internal occupational doses are specified in 10 CFR 20.1502, Conditions 22 Requiring Individual Monitoring of External and Internal Occupational Dose. Monitoring the intake of 23 radioactive material and assessing the committed effective dose equivalent (CEDE) (for internal exposures) is 24 required by 10 CFR 20.1502(b). The calculations licensees are required to perform in order to comply with 25 these regulations were affected by the 2007 revision of 10 CFR Part 20, Section 20.1003, Definitions and 26 10 CFR 50, Section 50.2, Definitions, (Ref. 2). This revision redefined the Total Effective Dose 27

DRAFT RG 8.34, Revision 1, Page 2 Equivalent (TEDE), as the sum of the effective dose equivalent (for external exposures) and the CEDE (for 28 internal exposures).

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

30 10 CFR 20.1007, Communications, 31 10 CFR 20.1202, Compliance with Requirements for Summation of External and Internal Doses, 32 10 CFR 20.1204, Determination of Internal Exposure, 33 10 CFR 20.1206, Planned Special Exposures, 34 10 CFR 20.1207, Occupational Dose Limits for Minors, 35 10 CFR 20.1208, Dose Equivalent to an Embryo/Fetus, 36 10 CFR 20.1501 Subpart F - Surveys and Monitoring (General) 37 10 CR 20.1703, Use of Individual Respiratory Protection Equipment, 38 10 CFR 20.2106, Records of Individual Monitoring Results, 39 10 CFR 20.2206, Reports of Individual Monitoring, and 40 10 CFR Part 19, Notices, Instructions, and Reports to Workers: Inspection and Investigations, 41 (Ref. 3).

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

46 Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation 47 Exposure Data, U.S. Nuclear Regulatory Commission, Washington, DC.

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

49 provides methods of determining intakes from bioassay results, 50 Regulatory Guide 8.11, Applications of Bioassay for Uranium 51 Regulatory Guide 8.25, Revision 1, Air Sampling in the Workplace (Ref. 5), provides 52 methods of determining intakes from air sampling measurements, 53 Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation 54 Exposure 55

DRAFT RG 8.34, Revision 1, Page 3 Regulatory Guide 8.35, Revision 1, "Planned Special Exposures (Ref. 6), provides guidance 56 on conducting planned special exposures, 57 Regulatory Guide 8.36, Radiation Dose to the Embryo/Fetus (Ref. 7), provides methods of 58 calculating doses to the embryo/fetus, and 59 Regulatory Guide 8.40, Methods for Measuring Effective Dose Equivalent from External 60 Exposure, (Ref. 8), provides details on acceptable methods of determining the effective 61 dose equivalent (from external exposure).

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

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

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

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

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

83 Determining the need for monitoring and demonstrating compliance, 84 Monitoring alpha intakes and determining internal dose, 85 Placement of dosimetry and resolving differences between passive and electronic 86 dosimeters, 87 Assessing intakes and committed dose equivalent from wounds, and 88

DRAFT RG 8.34, Revision 1, Page 4 Additional calculational methods of determining internal doses.

89

Background

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

98 Old definition:

TEDE = DDE + CEDE 99 New definition:

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

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

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

113 For protection against stochastic effects, the annual TEDE limit of is 5 rem (50 114 millisieverts (mSv)).

115 For protection of adults against nonstochastic effects, the annual total organ dose 116 equivalent (TODE) limit is 50 rem (500 mSv).

117 For protection of the lens of the eye, the annual lens dose equivalent (LDE) limit is 118 15 rem (150 mSv).

119 For protection of the skin of the whole body or to the skin of any extremity, the annual 120 shallow-dose equivalent (SDE) limit of 50 rem (500 mSv).

121

DRAFT RG 8.34, Revision 1, Page 5 For minors, 10 CFR 20.1207, Occupational Dose Limits for Minors, establishes an annual limit 122 at 10 percent of the adult limits.

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

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

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

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

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

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

DRAFT RG 8.34, Revision 1, Page 6 the TEDE ALARA, to increase monitoring (e.g., perform air sampling and Derived Air Concentrations 147 (DAC)-hour tracking and bioassay) and to limit intakes by the use of access controls, limiting exposures 148 times, or use respiratory protection equipment.

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

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

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

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

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

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

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

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

183

DRAFT RG 8.34, Revision 1, Page 7 The extent of the radiological characterization that is needed depends on the relative significance of 184 the alpha source term compared to other radiological contaminants. A site-specific alpha source term may be 185 used to identify radionuclides and determined their relative concentrations in a mixture, such as to comply 186 with the requirements of 10 CFR 20.1204(f). Once the relative concentrations are known, an effective 187 Derived Air Concentration (DAC) may be determined and used in radiological protection and dose 188 assessment.

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

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

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

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

201 Knowledge of the specific alpha radionuclide mix; 202 Conservatively assuming the most restrictive radionuclide in the mixture; or 203 Determining site specific, effective-DAC alpha values.

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

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

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

DRAFT RG 8.34, Revision 1, Page 8 encourages licensees to consult these international documents noted throughout this guide and implement the 220 good practices, where applicable that are consistent with NRC regulations. These documents are:

221 ICRP Publication 26, Recommendations of the International Commission on 222 Radiological Protection, (Ref. 14),

223 ICRP Publication 30, (7-volume set including supplements), Limits for Intakes of 224 Radionuclides by Workers, (Ref. 15),

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

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

240 C. STAFF REGULATORY GUIDANCE 241

1.

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

DRAFT RG 8.34, Revision 1, Page 9 and unlicensed2 sources under the control of the licensee. Monitoring is also required for any individual if the 247 external occupational dose is likely to exceed:

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

249 For minors in one year, a deep-dose equivalent of 0.1 rem (1 mSv), a lens dose equivalent 250 of 0.15 rem (1.5 mSv), and a shallow-dose equivalent to the skin of the whole body or to 251 the skin of the extremities of 0.5 rem (5 mSv).

252 For declared pregnant women during the entire pregnancy, a deep-dose equivalent of 0.1 253 rem (1 mSv).

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

256 For adults, 10 percent of the applicable annual limit on intake (ALI) 257 For minors in one year, 0.1 rem (1 mSv).

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

259

2.

Occupational Dose 260 The definition of occupational dose, in 10 CFR 20.1003, Definitions, includes dose received 261 during the course of employment in which assigned duties involve exposure to radiation or radioactive 262 material from licensed and unlicensed sources of radiation, whether in the possession of the licensee or other 263 person. The definition of occupational dose was changed in 1995 (60 FR 36038) (Ref. 19) such that 264 occupational dose applies to workers whose assigned duties involve exposure to radiation, irrespective of 265 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 10 their location inside or outside a restricted area. Note: A member of the public does not become an 266 occupationally exposed individual as a result of just entering a restricted area.

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

270

3.

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

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

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

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

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

295

4.

Determination of External Doses 296

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

DRAFT RG 8.34, Revision 1, Page 11 monitored, the CEDE can be assumed equal to zero, and the TEDE is equal to the EDEX. Similarly, if the 303 external dose is not monitored, the EDEX can be assumed equal to zero, and the TEDE is equal to the CEDE.

304

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

307 Measuring the DDE at the highest exposed part of the whole body with an external 308 personal monitoring device, as required by 10 CFR 20.1201(c), when an NRC method 309 for determining EDEX is not used.

310 Measuring external exposure with one or more external personal monitoring devices and 311 determining EDEX using an NRC approved method such as those provided in 312 Regulatory Guide 8.40, or as specifically approved by the NRC.

313 Calculating the EDEX based on survey data obtained under 10 CFR 20.1501 or other 314 radiological data, such as known source activity, dose rates, and exposure times using 315 scientifically sound technical methods. This may be required under unique exposure 316 situations (e.g., partial body exposed to radiation streaming of narrow beam geometries) 317 or when the individual monitoring device was not in the region of the highest whole body 318 exposure (per 10 CFR 20.1201(c)), or the results of the individual monitoring are not 319 available (i.e., damaged or lost device).

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

325

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

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

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

DRAFT RG 8.34, Revision 1, Page 12 possibly the dose rate). Due to the differences in dosimeter design and detection technology (and the 339 relative measurement errors associated with each) there can be differences in reading of these two 340 dosimeters for the same exposure, even if the dosimeters are co-located on the monitored individual.

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

346

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

352

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

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

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

367

5.

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

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

DRAFT RG 8.34, Revision 1, Page 13 8.25, Air Sampling in the Workplace and Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, 376 and Assumptions for a Bioassay Program for further guidance on determining uptakes and intakes.

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

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

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

388

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

391 I = C Air sample (µCi/ml)

• breathing rate (ml/minutes)

• exposure time (minutes);

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

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

397

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

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

410

DRAFT RG 8.34, Revision 1, Page 14 Note: The amount of the intake may be assessed using newer, updated biokinetic models 411 (e.g., ICRP Publications 60, 1990 Recommendations of the International Commission on 412 Radiological Protection, and ICRP Publication 103, The 2007 Recommendations of the 413 International Commission on Radiological Protection). However, the CEDE must be calculated 414 using the existing 10 CFR 20.1003 organ weighting factors (unless the use of other weighting factors 415 have been specifically approved by the NRC). In other words, the use of more recent tissue or organ 416 dose weighting factors is not acceptable (since the regulations in 10 CFR Part 20 list the specific 417 organ dose weighting factors that must be used).

418

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

424 Internal doses may be determined based on whole body count data and scaling factors when 425 nominal alpha doses occur, such as less than 500 mrem CEDE. However, if an alpha intake 426 exceeding a nominal level is considered likely, excreta sampling or lung counting may be needed to 427 determine intakes and assign dose. When excreta sampling is to be initiated, sampling should begin 428 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 429 least one sample is collected (following the first void for urine). ANSI N13.39 (2011), Design of 430 Internal Dosimetry Programs provides additional guidance on excreta sampling.

431

6.

Determination of Internal Dose 432

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

436

DRAFT RG 8.34, Revision 1, Page 15 Using dose coefficients3 from the U.S. Environmental Protection Agencys Federal 437 Guidance Report No. 11 (FGR-11) (Ref. 25).

438 Using ALI methods.

439 Using DAC-hour methods.

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

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

445

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

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

449

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

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

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

458 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 16 If both internal and external monitoring are being performed, the licensee must demonstrate 459 that both the 5 rem TEDE and the 50 rem TODE limits are met. One method of demonstrating 460 compliance with the TODE limit is by summing the DDE and the CDE to the highest exposed organ.

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

466

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

469 10 CFR 20.1201 also specifies two annual dose limits:

470 TODE limits (Section 20.1201(a)(1)(ii)) - the sum of the DDE and the CDE to any 471 individual organ or tissue other than the lens of the eye) being equal to 50 rem (0.5 Sv)),

472 and 473 SDE limits (Section 20.1201(a)(2)(ii)) - the SDE of 50 rem (0.5 Sv) to the skin of the 474 whole body or skin of any extremity.

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

481 In making the TODE dose calculation (to the skin organ) under 20.1201(a)(1)(ii), the DDE 482 component is zero, since for intakes by wounds, the DDE is zero (since DDE is an external 483 whole-body exposure). As a result, the calculated dose is only the CDE to the skin calculated to the 484 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 17 highest exposed, contiguous 10 cm2 area at a depth of 0.007 cm (in a manner similar to SDE 485 calculations).

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

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

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

494 30).

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

502

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

509

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

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

DRAFT RG 8.34, Revision 1, Page 18 accidental large exposures if more accurate information would lead to a better estimate of the actual 523 dose.

524

8.

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

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

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

539 540 541 542 543

DRAFT RG 8.34, Revision 1, Page 19 REFERENCES 544

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

546

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

547

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

549

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

551

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

553

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

555

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

557

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

559

9. Federal Register Notice 72 FR 68043 Occupational Dose Records, Labeling Containers, and the Total 560 Effective Dose Equivalent, U.S. Nuclear Regulatory Commission, Washington, DC, December 2007.5 561

10. Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure 562 Data U.S. Nuclear Regulatory Commission, Washington, DC 20555.

563

11. EPRI Alpha Monitoring and Control Guidelines for Operating Nuclear Power Stations, Revision 2, 564 August, 2013 (ML14083A535) (Ref.11) 565 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 20

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

567

13. International Commission on Radiological Protection (ICRP) Publication 2, Permissible Dose for 568 Internal Radiation, International Commission on Radiological Protection, Pergamon Press, Oxford, UK, 569 1959 570

14. ICRP Publication 26, Recommendations of the International Commission on Radiological Protection, 571 Protection, Pergamon Press, Oxford, UK, 1977.

572

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

574

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

576

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

578

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

580

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

582

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

21. IAEA Safety Standard Series No. RS-G-1.3, Assessment of Occupational Exposure Due To External 585 Sources of Radiation,, International Atomic Safety Agency, Vienna, Austria, 1999.

586

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

588

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

590

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

592

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

593

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

594 4.3, 1989.

595

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

596

DRAFT RG 8.34, Revision 1, Page 21

28. NCRP Report 87, NCRP Report No. 87, Use of Bioassay Procedures for Assessment of Internal 597 Radionuclide Deposition, Chapters 5.3.1, 5.3.2, 5.4.6 598

29. Toohey, R.E., et al, "Dose Coefficients for Intakes of Radionuclides Via Contaminated Wounds, Health 599 Phys. 2011 May;100(5):508-14; also available from the Oak Ridge Institute for Science and Education at 600 URL: http://orise.orau.gov/reacts/resources/retention-intake-publication.aspx 601

30. Ishique, N., "Implementation of the NCRP Wound Model for Interpretation of Bioassay Data for Intake 602 of Radionuclides Through Contaminated Wounds", Radiation Research. 2009 May; 50(3): 267-276 603

31. NCRP Report No. 156, Development of a Biokinetic Model for Radionuclide-Contaminated Wounds 604 and Procedures for Their Assessment, Dosimetry, and Treatment National Council on Radiation 605 Protection and Measurements, Library of Congress, Washington, DC.

606

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

608

33. ANSI N13.39 (2011), Design of Internal Dosimetry Programs 609

34. Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay 610 Program, U.S. Nuclear Regulatory Commission, Washington, DC.

611

35. Regulatory Guide 8.11, Applications of Bioassay for Uranium 612

36. Regulatory Guide 8.29, Instruction Concerning Risks from Occupational Radiation Exposure 613

37. Regulatory Guide 8.7, Instructions for Recording and Reporting Occupational Radiation Exposure 614 Data, U.S. Nuclear Regulatory Commission, Washington, DC.

615

38. K. F. Eckerman, A. B. Wolbarst, and A. C. B. Richardson, Limiting Values of Radionuclide Intake and 616 Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal 617 Guidance Report No. 11 (EPA 520/1-8-020), U.S. Environmental Protection Agency, Washington, DC, 618 1988 619

39. Regulatory Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay 620 Program, Chapter 3, page 8.9-4 621

40. Information Notice No. 97-36: (June 27,1997), Unplanned Intakes by Worker of Transuranic Airborne 622 Radioactive Materials and External Exposure due to Inadequate Control of Work 623

41. SECY-98-245 - Rulemaking Plan - Protection Against Discrete Radioactive Particle" 624

42. NUREG/CR-6918, VARSKIN 4: A Computer Code for Assessing Skin Dose from Skin 625 Contamination 626

43. NCRP Report No. 106, Limit for Exposure to Hot Particles on the Skin (1980) 627 628

DRAFT RG 8.34, Revision 1, Page 22 Appendix A 629 630

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

636 637 (1)

DCs for the CDE to an organ or tissue per unit of activity (DCorgan) (e.g. the heading 638 Lung below) and 639 640 (2)

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

648 649 CDE (rem) = DCorgan (rem/µCi)

• I (µCi) 650 CEDE (rem) = DCeffective (rem/Ci)

• I (Ci) 651 652 Example 1:

Calculations of the CDE and the CEDE for Co60, based on bioassay measurements 653 using the DCs from FGR-11.

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

656

DRAFT RG 8.34, Revision 1, Page 23 From Table 2.1 of FGR-11 (see excerpt below), the DCs for Class Y, Co-60 radionuclide are 657 3.45E-7 Sv/Bq for the CDE and 5.91E-8 Sv/Bq for the CEDE.

658 659 Excerpt from Federal Guidance Report No. 11 660 661 The DCs in FGR-11 are tabulated in Sv/Bq and may be converted to mrem/Ci by 662 multiplying by 3.7x109.

663 664 DClung = (3.45E-7 Sv/Bq) * (3.7E+9) = 1,277 mrem/µCi 665 DCeffective = (5.91E-8 Sv/Bq) * (3.7E+9) = 219 mrem/µCi 666 667 The doses are calculated by multiplying these DCs by the intake of 0.36 µCi.

668 669 CDElung = (1,277 mrem/µCi) * (0.36 µCi) = 460 mrem 670 CEDE = (219 mrem/µCi) * (0.36 µCi) = 79 mrem 671 672

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

678 679 680 681

DRAFT RG 8.34, Revision 1, Page 24 Since the stochastic ALI corresponds to a 5 rem (50 mSv) CEDE dose limit, the CEDE may 682 be calculated based on the ratio of the intake to the stochastic ALI, multiplied by 5 rem (50 mSv).

683 684 CEDE = (I/ALI)

• 5 rem 685 686 Example 2:

Calculate the CEDE based on bioassay measurements using the stochastic ALIs.

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

689 690 From Appendix B above, Table 1, Column 2, the ALI for Class Y Co-60 is:

691 692 ALI (stochastic) = 30 µCi 693 CEDE = (I/ALI)

• 5 rem 694 CEDE = (0.36 µCi/30 µCi)

• 5 rem = 0.06 rem = 60 mrem 695 696 Note: Considering the precision of a 1 significant figure for the ALI values, this 60 mrem 697 value compares favorably to the calculated CEDE value of 79 mrem determined in Example 698 A above using the FGR-11 method. Either calculational method and/or result is acceptable in 699 demonstrating compliance.

700 701

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

706 9E+1 µCi for ingestion and 2E+2 µCi for inhalation in excerpt below).

707 708 709 710 711

DRAFT RG 8.34, Revision 1, Page 25 Since the nonstochastic ALI corresponds to a 50 rem (500 mSv) CDE dose limit, the CDE 712 may be calculated based on the ratio of the intake to the nonstochastic ALI, multiplied by 50 rem (500 713 mSv).

714 CDE = (I/ALI)

• 50 rem 715 716 Note: For a mixture of radionuclides, the sum of the fractions technique as described in 10 CFR 717 20.1202(b) must be used.

718 Example 3:

Calculate the CDE based on bioassay measurements using the nonstochastic ALIs.

719 720 The intake by inhalation for a worker was estimated by bioassay to be 131 nCi (0.131 µCi) of 721 I-131, Class D aerosol. Calculate the CDE to the thyroid.

722 723 From Appendix B above, Table 1, Column 2, the ALI for Class D I-131 is:

724 725 ALI (nonstochastic) = 5E+1 µCi = 50 µCi 726 CDE = (0.131 µCi/50 µCi)

• 50 rem = 0.131 rem = 131 mrem 727 728

4. Calculation of the CDE based on air sampling and nonstochastic DAC-hrs 729 For nonstochastic radionuclides, an exposure to an airborne concentration of 1 DAC results 730 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 731 per DAC-hour.

732 CDE = [25 mrem per DAC-hr]

• number of DAC-hrs 733 where the number of DAC-hrs = (air concentration / DAC value)

• exposure time 734 Example 4:

Calculation of the CDE based on air sampling and nonstochastic DAC-hrs.

735 736 Calculate the CDE to the thyroid for a 30-minute exposure based on an air sample result of 737 2.1E-7 µCi/ml (I-131).

738 739 The nonstochastic DAC for I-131 is listed in Appendix B (see excerpt below) as 2E-8 740

µCi/ml.

741 742

DRAFT RG 8.34, Revision 1, Page 26 743 744 CDE = 25 mrem/DAC-hr * [(2.1E-7 µCi/ml) / (2E-8 µCi/ml)] number of DACs * (0.5 hrs) = 131 745 mrem 746

5. Calculations of the CEDE based on air sampling and stochastic DAC-hrs 747 For stochastic radionuclides (e.g., Co-60), an exposure to an airborne concentration of 1 DAC results 748 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 749 CEDE per stochastic DAC-hr.

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

• No. of DAC-hrs 751 where the number of DAC-hrs = (air concentration / DAC value)

• exposure time 752 Example 5:

Calculation of the CEDE based on air sampling and stochastic DAC-hrs.

753 754 Calculate the CEDE for a 30-minute exposure based on an air sample result of 2.1E-7 µCi/ml 755 (Co-60).

756 757 From Appendix B below, the stochastic DAC for Co-60 clearance Class Y compound is 1E-8 758

µCi/ml.

759 760 761 CEDE = [2.5 mrem/DAC-hr] * [(2.1E-7 µCi/ml) / (1E-8 µCi/ml)] No. of DACs* (0.5 hrs) = 26 mrem 762

DRAFT RG 8.34, Revision 1, Page 27

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

• No. of DAC-hrs 764 No. DAC-hrs = [air concentration / calculated DAC value] * [exposure time]

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

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

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

770 771 Example 6:

Calculation of the CEDE based on air sampling and calculated stochastic DAC-hrs.

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

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

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