ML21197A156

Rev.3 - ACRS pre-decisional Version

ML21197A156
Person / Time
Issue date:	07/16/2021
From:	Steven Garry NRC/NRR/DRA/ARCB
To:
Song K
Shared Package
ML21132A170	List: ML21132A226 ML21132A291 ML21197A156 ML21197A158 Regulatory Guide 1.21
References
DG 1377 RG 1.21 Rev 3
Download: ML21197A156 (85)

Text

U.S. NUCLEAR REGULATORY COMMISSION REGULATORY GUIDE RG 1.21 Issue Date: Month 2021 Technical Lead: Steven Garry 1 MEASURING, EVALUATING, AND REPORTING 2 RADIOACTIVE MATERIAL IN LIQUID AND GASEOUS 3 EFFLUENTS AND SOLID WASTE 4

5 A. INTRODUCTION 6

7 Purpose 8

9 This regulatory guide (RG) describes methods the staff of the U.S. Nuclear Regulatory 10 Commission (NRC) considers acceptable for the following uses:

11 12 (1) measuring, evaluating, and reporting licensed (plant-related) radioactivity in effluents and 13 solid radioactive waste shipments from nuclear power plants and spent fuel storage facilities, 14 and 15 16 (2) assessing and reporting the public dose to demonstrate compliance with Title 10 of the Code 17 of Federal Regulations (10 CFR) Part 20, Standards for Protection Against Radiation 18 (Ref. 1), Title 40, (40 CFR) Part 190, Environmental Radiation Protection Standards for 19 Nuclear Power Operations (Ref. 2), and nuclear power plant Technical Specifications.

20 21 This guide incorporates the risk-informed principles of the Reactor Oversight Process. A 22 risk-informed, performance-based approach to regulatory decision making combines the risk-informed 23 and performance-based elements discussed in the staff requirements memorandum to SECY-98-144, 24 Staff RequirementsSECY-98-144White Paper on Risk-Informed and Performance-Based 25 Regulation, dated February 24, 1999 (Ref. 3).

26 27 Applicability 28 29 This RG is a Division 1, Power Reactors RG, which applies to nuclear power plant licensees 30 and applicants subject to Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for 31 Protection Against Radiation, This RG is also applicable to specific and general licensees under 10 Part 32 72 for storage of spent fuel.

33 This includes licenses issued under the following regulations:

34 Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site in the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/,under Document Collections, in Regulatory Guides, at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/contactus.html.

Electronic copies of this RG, previous versions of RGs, and other recently issued guides are also available through the NRCs public Web site in the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/, under Document Collections, in Regulatory Guides. This RG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/readingrm/adams.html, under ADAMS Accession Number (No.)

ML21133A019. The regulatory analysis may be found in ADAMS under Accession No. ML20287A434. The associated draft guide DG-1377 may be found in ADAMS under Accession No. ML20287A423, and the staff responses to the public comments on DG-1377 may be found under ADAMS Accession No. ML21132A226.

35

• 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 4), applies to 36 the licensing of production and utilization facilities.

37 38

• 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 5),

39 applies to applicants and holders of combined licenses, standard design certifications, standard 40 design approvals, and manufacturing licenses.

41 42

• 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 43 High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste (Ref. 6),

44 applies to general licenses issued under Part 72 and to applicants for and holders of specific 45 licenses under Part 72.

46 47 Applicable Regulations 48 49 The following regulations establish the regulatory basis for the radiological effluent control 50 program:

51 52

• 10 CFR Part 20, Standards for Protection Against Radiation 53 54 o 10 CFR 20.1003, Definitions, defines terminology that is used in the regulations and in 55 this regulatory guide.

56 57 o 10 CFR 20.1301, Dose limits for individual members of the public, establishes 58 radiation dose limits for individual members of the public.

59 60 o 10 CFR 20.1302, Compliance with dose limits for individual members of the public, 61 requires licensees to perform surveys of radiation levels in unrestricted and controlled 62 areas and radioactive materials in effluents released to unrestricted and controlled areas to 63 demonstrate compliance with the dose limits for individual members of the public.

64 65 o 10 CFR 20.1402, Radiological criteria for unrestricted use, establishes acceptance 66 criteria for license termination to achieve the sites unrestricted use status after 67 decommissioning.

68 69 o 10 CFR 20.1501, General, establishes requirements for performing radiological 70 surveys.

71 72 o 10 CFR 20.2001, General requirements (for waste disposal), establishes methods for 73 disposing of licensed material.

74 75 o 10 CFR 20.2103, Records of surveys, requires licensees to maintain records of surveys 76 and calibrations.

77 78 o 10 CFR 20.2107, Records of dose to individual members of the public, requires 79 licensees to maintain records that demonstrate compliance with dose limits for members 80 of the public.

81 82 o 10 CFR 20.2108, Records of waste disposal, requires licensees to maintain records of 83 the disposal of licensed material.

84 RG 1.21, Rev. 3, Page 2

85 o 10 CFR Part 20, Appendix B, Annual Limits on Intakes (ALIs) and Derived Air 86 Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent 87 Concentrations; Concentrations for Release to Sewage, establishes intake limits and 88 airborne and liquid concentration limits for occupational exposure and member of the 89 public exposure.

90 91

• 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities 92 93 o 10 CFR 50.34a, Design objectives for equipment to control releases of radioactive 94 material in effluentsnuclear power reactors, establishes numerical guides for design 95 objectives and limiting conditions of operation to control radioactive effluents.

96 97 o 10 CFR 50.36a, Technical specifications on effluents from nuclear power reactors, 98 requires licensees to establish technical specifications with operating procedures and 99 controls be established and followed and that the radioactive waste system be maintained 100 and used.

101 102 o 10 CFR 50.75, Reporting and record keeping for decommissioning planning, 103 paragraph (g,) requires licensees to keep records of information important to 104 decommissioning.

105 106 o 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, 107 General Design Criterion (GDC) 60, Control of Releases of Radioactive Materials to the 108 Environment, specifies that the nuclear power unit design shall include means to control 109 suitably liquid and gaseous effluents and solid waste.

110 111 o 10 CFR Part 50, Appendix A, GDC 64, Monitoring Radioactivity Releases, specifies 112 that means shall be provided for monitoring the reactor containment atmosphere, spaces 113 containing components for recirculation of loss-of-coolant fluids, effluent discharge paths 114 and the plant environs for radioactivity that may be released from normal operations, 115 anticipated operational occurrences, and from postulated accidents.

116 117 o 10 CFR Part 50, Appendix I, Numerical Guides for Design Objectives and Limiting 118 Conditions for Operation to Meet the Criterion As Low As Is Reasonably Achievable 119 (ALARA) for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor 120 Effluents, establishes design objectives for meeting the requirements of 10 CFR 50.34a.

121 122

• 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants 123 124 o 10 CFR 52.0, Scope, requires Part 52 licensees to comply with all requirements in 125 10 CFR Chapter I that are applicable, which includes, for example, 10 CFR Part 20 as 126 discussed above.

127 128

• 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 129 High Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste 130 131 o 10 CFR 72.44(d) requires that each specific license must include technical specifications 132 that establishes limits on the release of radioactive materials and the ALARA objectives 133 for effluents and that require establishment of an environmental monitoring program to 134 ensure compliance with those limits RG 1.21, Rev. 3, Page 3

135 136 o 10 CFR 72.104, Criteria for radioactive materials in effluents and direct radiation from 137 an ISFSI or MRS, establishes dose limits to any real individual (excluding occupational 138 exposures) beyond the Part 72 controlled area (as defined in 10 CFR 72.3 and meeting 139 the minimum size requirements in 72.106(b))

140 141 o 10 CFR 72.126, Criteria for radiological protection, requires radiation protection 142 systems be provided with effluent and direct radiation monitoring systems and controls to 143 limit releases to ALARA under normal conditions and control releases under accident 144 conditions and ensure limits relating to releases to the general environment will not be 145 exceeded 146 147

• 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations 148 149 o 40 CFR 190.10, Standards for normal operation, establishes standards for normal 150 operations and annual dose equivalent standards and limits on the total quantity of 151 radioactive materials entering the environment from the entire uranium fuel cycle.

152 153 o 40 CFR 190.11, Variances for unusual operations, establishes variances (allowances) 154 for unusual operations where the standards in 40 CFR 190.10 may be exceeded.

155 156

• 40 CFR Part 191, Environmental Radiation Protection Standards for Management and Disposal 157 of Spent Nuclear Fuel and Transuranic Radioactive Wastes (Ref. 7) 158 159 o 40 CFR 191.03(a), Standards, establishes standards for the management and storage of 160 spent nuclear fuel or transuranic radioactive wastes.

161 162 163 Related Guidance 164 165

• RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants (Ref. 8), provides 166 guidance for an onsite meteorological measurements program.

167 168

• RG 1.97, Revisions 0, 1, 2 and 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants 169 to Assess Plant and Environs Conditions During and Following an Accident, issued 170 December 1975, August 1977, and December 1980, and May 1983, respectively (Ref. 9),

171 provides guidance on instrumentation used to monitor plant variables and systems during and 172 following an accident.

173 174

• RG 1.97, Revision 4, Criteria for Accident Monitoring Instrumentation for Nuclear Power 175 Plants, issued June 2006 (Ref. 10), endorses (with certain clarifying regulatory positions) the 176 Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 497-2002, IEEE 177 Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating 178 Stations (Ref. 11).

179 180

• RG 1.97, Revision 5, Criteria for Accident Monitoring Instrumentation for Nuclear Power 181 Plants, issued April 2019 (Ref. 12), endorses, with exceptions and clarifications, IEEE 182 Std. 497-2016, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear 183 Power Generating Stations (Ref. 13).

184 RG 1.21, Rev. 3, Page 4

185

• RG 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for 186 the Purpose of Demonstrating Compliance with 10 CFR Part 50, Appendix I (Ref. 14), describes 187 basic features of calculational models and assumptions used for the estimation of doses to the 188 public.

189 190

• RG 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents 191 in Routine Releases from Light-Water-Cooled Reactors (Ref. 15), describes models and 192 assumptions for the estimation of atmospheric dispersion of gaseous effluent releases.

193 194

• RG 1.112, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents 195 from Light-Water-Cooled Power Reactors, (Ref. 16), provides acceptable methods for applicants 196 to construct a nuclear power reactor to calculate realistic radioactive source terms for use in 197 evaluating radioactive waste treatment systems to determine whether the design objectives of 198 10 CFR Part 50, Appendix I, are met, and to assess the environmental impact of radioactive 199 effluents.

200 201

• RG 1.113, Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor 202 Releases for the Purpose of Implementing Appendix I (Ref. 17), describes general approaches 203 for the analysis of releases of liquid effluents into surface water bodies.

204 205

• RG 1.184, "Decommissioning of Nuclear Power Reactors" (Ref. 18), provides guidance that 206 during decommissioning, Technical Specifications require operational procedures for the control 207 of effluent releases and submittal of annual effluent reports as specified by 10 CFR 50.36a.

208 209

• RG 1.185, Standard Format and Content for Post-Shutdown Decommissioning Activities 210 Report (Ref. 19), identifies information licensees should provide to NRC and the public of the 211 licensees expected decommissioning activities and schedule.

212 213

• RG 4.1, Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants 214 (Ref. 20), describes acceptable programs for establishing and conducting an environmental 215 monitoring program.

216 217

• RG 4.13, Environmental DosimetryPerformance Specifications, Testing, and Data Analysis 218 (Ref. 21), provides specifications for environmental dosimetry and methods of analyzing 219 dosimetry to determine dose to members of the public.

220 221

• RG 4.15, Quality Assurance for Radiological Monitoring Programs (Inception Through Normal 222 Operations to License Termination)Effluent Streams and the Environment (Ref. 22), describes 223 design and implementation programs to ensure the quality of the results of measurements of 224 radioactive materials in the effluents from, and environment outside of, facilities that process, 225 use, or store radioactive materials.

226 227

• RG 4.20, Constraint on Releases of Airborne Radioactive Materials to the Environment for 228 Licensees other than Power Reactors (Ref. 23), provides guidance for meeting the constraint on 229 airborne emissions of radioactive material as described in 10 CFR 20.1101(d).

230 231

• RG 4.25, Assessment of Abnormal Radionuclide Discharges in Groundwater to the Unrestricted 232 Area at Nuclear Power Plant Sites (Ref. 24), describes an approach that is acceptable for use in 233 assessing abnormal discharges of radionuclides in groundwater from the subsurface to the 234 unrestricted area at nuclear power plant sites.

RG 1.21, Rev. 3, Page 5

235 236

• Generic Letter (GL) 89-01, Guidance for the Implementation of Programmatic Controls for 237 Radiological Effluent Technical Specifications in the Administrative Controls Section of 238 Technical Specifications and the Relocation of Procedural Details to the Offsite Dose Calculation 239 Manual or Process Control Program, dated January 31, 1989 (Ref. 25), provides guidance for the 240 preparation of a license amendment request to relocate programmatic controls for radioactive 241 effluents and for radiological environmental monitoring from technical specifications to the 242 licensee-controlled Offsite Dose Calculation Manual (ODCM) or equivalent document.

243 244

• NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power 245 Plants" issued October 1978 (Ref. 26), is one of the bases documents for the Radioactive Effluent 246 Controls Program in Standard Technical Specifications (section 5.5.4).

247 248

• NUREG-0016, Revision 1 and Revision 2, Calculation of Releases of Radioactive Materials in 249 Gaseous and Liquid Effluents from Boiling-Water Reactors (GALE-BWR 3.2 Code), issued 250 January 1979 and July 2020, respectively (Ref. 27), is a computerized mathematical model for 251 calculating the release of radioactive materials in gaseous and liquid effluents from boiling-water 252 reactors (BWRs).

253 254

• NUREG-0017, Revision 1 and Revision 2, Calculation of Releases of Radioactive Materials in 255 Gaseous and Liquid Effluents from Pressurized-Water Reactors (GALE-PWR 3.2 Code), issued 256 April 1985 and July 2020, respectively (Ref. 28), is a computerized mathematical model for 257 calculating the release of radioactive materials in gaseous and liquid effluents from 258 pressurized-water reactors (PWRs).

259 260

• NUREG-0543, Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel 261 Cycle Standard (CFR Part 190), 1980 (Ref. 29) explains the rationale for using Appendix I to 262 demonstrate compliance with 40 CFR 190 and methods for demonstrating compliance when 263 radioactive effluents exceed Appendix I numerical guidance.

264 265

• NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980 266 (Ref. 30), provides specific items that were approved by the NRC Commission following the 267 accident at Three Mile Island Nuclear Station (TMI) for implementation at reactors.

268 269

• NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 270 Controls for Pressurized Water Reactors, issued April 1991 (Ref. 31), provides the PWR effluent 271 controls that may be removed from technical specifications and incorporated into the licensees 272 ODCM (or equivalent).

273 274

• NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 275 Controls for Boiling Water Reactors, issued April 1991 (Ref. 32), provides the BWR effluent 276 controls that may be removed from technical specifications and incorporated into the licensees 277 ODCM (or equivalent).

278 279

• NUREG-1575, Rev. 1, Multi-Agency Radiation Survey and Site Investigation Manual 280 (MARSSIM) (Ref. 33), provides information on planning, conducting, evaluating, and 281 documenting building surface and surface soil final status radiological surveys for demonstrating 282 compliance with dose or risk-based regulations or standards.

283 RG 1.21, Rev. 3, Page 6

284

• NUREG-1576, Multi-Agency Radiological Laboratory Analytical Protocols Manual (Ref. 34),

285 provides guidance for the planning, implementation, and assessment of projects that require the 286 laboratory analysis of radionuclides.

287 288

• NUREG-1757, Volume 2, Revision 1, Consolidated Decommissioning Guidance:

289 Characterization, Survey, and Determination of Radiological Criteria (Ref. 35), provides 290 guidance on compliance with 10 CFR Part 20, Subpart E - Radiological Criteria for License 291 Termination.

292 293

• NUREG-1940, RASCAL 4: Description of Models and Methods, issued December 2012 294 (Ref. 36), provides a description of an emergency response consequence assessment tool 295 including models and methods for source term calculations, atmospheric dispersion and 296 deposition, and dose calculations.

297 298

• NUREG-1940, Supplement 1, RASCAL 4.3: Description of Models and Methods, issued 299 May 2015 (Ref. 37), describes the Radiological Assessment System for Consequence Analysis 300 (RASCAL) models and methods for source term calculations, atmospheric dispersion and 301 deposition, and dose calculations for accident analysis.

302 303

• NUREG/CR-6948, Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities 304 and Sites issued November 2007 (Ref. 38), presents a framework for assessing what, where, 305 when, and how to monitor contamination in groundwater.

306 307

• NUREG/CR-6805, A Comprehensive Strategy of Hydrogeology Modeling and Uncertainty 308 Analysis for Nuclear Facilities and Sites, issued July 2003 (Ref. 39), describes a strategy for a 309 systematic and comprehensive approach to hydrogeologic conceptualization, model development, 310 and predictive uncertainty analysis.

311 312 Purpose of Regulatory Guides 313 314 The NRC issues RGs to describe methods that are acceptable to the staff for implementing 315 specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific 316 issues or postulated events, and to describe information that the staff needs in its review of applications 317 for permits and licenses. RGs are not NRC regulations and compliance with them is not required.

318 Methods and solutions that differ from those set forth in RGs are acceptable if supported by a basis for the 319 issuance or continuance of a permit or license by the Commission.

320 321 Paperwork Reduction Act 322 323 This RG provides voluntary guidance for implementing the mandatory information collections in 324 10 CFR Parts 20, 50, 52, 72, that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et.

325 seq.). These information collections were approved by the Office of Management and Budget (OMB),

326 approval numbers 3150-0014, 3150-0011, 3150-0151, and 3150-0132, respectively. Send comments 327 regarding this information collection to the FOIA, Library, and Information Collections Branch (T6-328 A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to 329 Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and 330 Regulatory Affairs (3150-0014, 3150-0011, 3150-0151, and 3150-0132), Attn: Desk Officer for the 331 Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC20503; e- mail:

332 oira_submission@omb.eop.gov.

333 RG 1.21, Rev. 3, Page 7

334 Public Protection Notification 335 336 The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of 337 information unless the document requesting or requiring the collection displays a currently valid OMB 338 control number.

339 RG 1.21, Rev. 3, Page 8

340 TABLE OF CONTENTS 341 342 A. INTRODUCTION .................................................................................................................................. 1 343 B. DISCUSSION ....................................................................................................................................... 11 344 Reason for Revision................................................................................................................................ 11 345 Background............................................................................................................................................. 11 346 Objectives of the Radiological Effluent Controls Program .................................................................... 12 347 C. STAFF REGULATORY GUIDANCE ................................................................................................. 15 348 1. Effluent Monitoring .................................................................................................................... 15 349 1.1 Effluent Monitoring Programs ................................................................................................ 15 350 1.2 Release Points for Effluent Monitoring .................................................................................. 15 351 1.3 Monitoring a Significant Release Point .................................................................................. 16 352 1.4 Monitoring a Less-Significant Release Point .......................................................................... 16 353 1.5 Monitoring Leaks and Spills ................................................................................................... 17 354 1.6 Monitoring Continuous Releases ............................................................................................ 19 355 1.7 Monitoring Batch Releases ..................................................................................................... 20 356 1.8 Principal Radionuclides for Effluent Monitoring ................................................................... 20 357 1.9 Carbon-14 ............................................................................................................................... 22 358 1.10 Return/Reuse of Previously Discharged Radioactive Effluents .............................................. 23 359 1.11 Abnormal Releases and Abnormal Discharges ....................................................................... 23 360 2. Effluent Sampling ....................................................................................................................... 24 361 2.1 Representative Sampling......................................................................................................... 24 362 2.2 Sampling Liquid Radioactive Waste ....................................................................................... 25 363 2.3 Sampling Gaseous Radioactive Waste .................................................................................... 25 364 2.4 Sampling Bias ......................................................................................................................... 25 365 2.5 Composite Sampling ............................................................................................................... 26 366 2.6 Sample Preparation and Preservation...................................................................................... 26 367 2.7 Short-Lived Radionuclides and Decay Corrections ................................................................ 26 368 3. Effluent Dispersion (Meteorology and Hydrology) .................................................................... 26 369 3.1 Meteorological Data ................................................................................................................ 26 370 3.2 Atmospheric Dispersion (Transport and Diffusion) ............................................................... 27 371 3.3 Release Height ........................................................................................................................ 27 372 3.4 Aquatic Dispersion (Surface Waters)...................................................................................... 28 373 3.5 Spills and Leaks to the Ground Surface .................................................................................. 28 374 3.6 Spills and Leaks to Groundwater ............................................................................................ 28 375 4. Quality Assurance ....................................................................................................................... 30 376 4.1 Quality Assurance Programs ................................................................................................... 30 377 4.2 Quality Control Checks ........................................................................................................... 31 378 4.3 Surveillance Frequencies ........................................................................................................ 31 379 4.4 Procedures ............................................................................................................................... 31 380 4.5 Calibration of Laboratory Equipment and Routine Effluent Radiation Monitors................... 31 381 4.6 Calibration of Measuring and Test Equipment ....................................................................... 32 382 4.7 Calibration Frequency ............................................................................................................. 32 383 4.8 Measurement Uncertainty ....................................................................................................... 32 384 4.9 Calibration of Accident-Range Radiation Monitors and Accident-Range Effluent Monitors 32 385 5. Dose Assessments for Individual Members of the Public .......................................................... 34 386 5.1 Bounding Assessments ........................................................................................................... 35 387 5.2 Individual Members of the Public ........................................................................................... 36 388 5.3 Occupancy Factors .................................................................................................................. 36 389 5.4 10 CFR Part 50, Appendix I.................................................................................................... 36 RG 1.21, Rev. 3, Page 9

390 5.5 10 CFR 20.1301(a) through (c) ............................................................................................... 37 391 5.6 10 CFR 20.1301(e).................................................................................................................. 37 392 5.7 Dose Assessments for 10 CFR Part 50, Appendix I ............................................................... 38 393 5.8 Dose Assessments for 10 CFR 20.1301(e) ............................................................................. 39 394 5.9 Dose Calculations ................................................................................................................... 40 395 6. Solid Radioactive Waste Released from the Unit ...................................................................... 40 396 7. Reporting Errata in Effluent Release Reports ............................................................................. 41 397 7.1 Examples of Small Errors ....................................................................................................... 41 398 7.2 Reporting Small Errors ........................................................................................................... 41 399 7.3 Examples of Large Errors ....................................................................................................... 41 400 7.4 Reporting Large Errors ........................................................................................................... 42 401 8. Changes to Effluent and Environmental Programs ..................................................................... 43 402 9. Format and Content of the Annual Radioactive Effluent Release Report ...................................... 43 403 9.1 Gaseous Effluents ................................................................................................................... 44 404 9.2 Liquid Effluents ...................................................................................................................... 46 405 9.3 Solid Waste Shipments Released from the Unit (per Standard Technical Specifications) ..... 47 406 9.4 Dose Assessments ................................................................................................................... 48 407 9.5 Supplemental Information....................................................................................................... 48 408 D. IMPLEMENTATION ........................................................................................................................... 52 409 GLOSSARY ............................................................................................................................................... 53 410 REFERENCES ........................................................................................................................................... 62 411 BIBLIOGRAPHY ....................................................................................................................................... 69 412 APPENDIX ATABLES ............................................................................................................................ 1 413 414 RG 1.21, Rev. 3, Page 10

415 B. DISCUSSION 416 417 Reason for Revision 418 419 This revision of RG 1.21 (Revision 3):

420

• Provides guidance and acceptable methods for calibration of accident range radiation monitors 421 and accident range effluent monitors, 422

• Revises guidance on recommendations for reviewing and updating long-term, annual average /Q 423 and D/Q values, 424

• Clarifies reporting requirements for low level radioactive waste (LLW) shipments, specifically 425 that the report includes the waste shipped from the unit (plant site), and that waste classification 426 does not need to be reported when shipped from the unit (plant site) to a waste processor, 427

• Clarifies the existing guidance in NUREG 1301 and NUREG 1302 that environmental monitoring 428 for iodine (I) -131 in drinking water should be performed if a prospective dose evaluation of the 429 annual thyroid dose from I-131 to a person in any age group from the drinking water route of 430 exposure is greater than one mrem.

431

• Clarifies the existing process as currently described in Technical Specifications for making 432 changes to effluent and environmental programs, and, 433

• Incorporates the existing Regulatory Issue Summary 2008-03, Return/Reuse of Previously 434 Discharged Radioactive Effluents (Ref. 40).

435 436 Background 437 438 In addition to this RG, five additional basic documents contain the primary regulatory guidance 439 for implementing the 10 CFR Part 20 and 10 CFR Part 50 regulatory requirements and plant technical 440 specifications related to monitoring and reporting of radioactive material in effluents and environmental 441 media, solid radioactive waste shipments, and the public dose that results from licensed operation of a 442 nuclear power plant:

443 444 (1) RG 4.1 445 (2) RG 4.15 446 (3) RG 1.109 447 (4) NUREG-1301 448 (5) NUREG-1302 449 450 These documents, when used in an integrated manner, provide the basic guidance and 451 implementation details for developing and maintaining effluent and environmental monitoring programs 452 at nuclear power plants. The four RGs (RG 1.21, RG 4.1, RG 4.15, and RG 1.109) specify the guidance 453 for radiological monitoring and the assessment of dose, and the two NUREGs (NUREG-1301 and 454 NUREG-1302) provide specific implementation details for effluent and environmental monitoring 455 programs.

456 457 RG 1.21 addresses the measuring, evaluating, and reporting of effluent releases, solid radioactive 458 waste shipments, and public dose from nuclear power plants. The guide describes the important concepts 459 in planning and implementing an effluent and solid radioactive waste program. Concepts covered include 460 meteorology, release points, monitoring methods, identification of principal radionuclides, unrestricted 461 area boundaries, continuous and batch release methods, representative sampling, composite sampling, 462 radioactivity measurements, decay corrections, quality assurance (QA), solid radioactive waste shipments, RG 1.21, Rev. 3, Page 11

463 and public dose assessments. The dose to occupational workers, including contributions from activities 464 associated with effluent programs (such as LLW processing, storage, and shipping, as well as dose from 465 handling resins and filters for gaseous and liquid radioactive waste), is occupational dose associated with 466 the licensed operation and is not included in RG 1.21.

467 468 RG 4.1 addresses the environmental monitoring program. The guide discusses principles and 469 concepts important to environmental monitoring at nuclear power plants. The RG provides guidance on 470 both the preoperational and operational Radiological Environmental Monitoring Programs for the 471 routinely monitored exposure pathways (inhalation, ingestion, and direct radiation). The guide defines 472 the sampling media and sampling frequency, and the methods of comparing environmental measurements 473 to effluent releases in the Annual Radiological Environmental Operating Report (AREOR).

474 475 RG 4.15 provides the basic principles of QA in all types of radiological monitoring programs for 476 effluent streams and the environment. The guide provides principles for structuring organizational lines 477 of communication and responsibility, using qualified personnel, implementing standard operating 478 procedures, defining data quality objectives (DQOs), performing quality control (QC) checking for 479 sampling and analysis, auditing the process, and taking corrective actions.

480 481 RG 1.109 provides the detailed implementation guidance for demonstrating that radioactive 482 effluents conform to ALARA design objectives of 10 CFR Part 50, Appendix I. The RG describes 483 calculational models and parameters for estimating dose from effluent releases, including the dispersion 484 of the effluent in the atmosphere and surface water bodies.

485 486 NUREG-1301 and NUREG-1302 provide the detailed implementation guidance by describing 487 effluent and environmental monitoring programs. These NUREGs provide guidance on meeting effluent 488 monitoring and environmental sampling requirements, surveillance requirements for effluent monitors, 489 types of monitors and samplers, sampling and analysis frequencies, types of analysis and radionuclides 490 analyzed, lower limits of detection (LLDs), specific environmental media to be sampled, and reporting 491 and program evaluation and revision.

492 493 Objectives of the Radiological Effluent Controls Program 494 495 The requirements for the radiological effluent control program are in 10 CFR Part 20 and the 496 technical specifications that are part of a license, including limitations on dose conforming to 497 10 CFR Part 50, Appendix I. In addition, a facilitys technical specifications describe specific regulatory 498 requirements. Licensees can use these regulatory requirements and the RG 1.21 regulatory guidance as a 499 basis for establishing the radiological effluent control program. The radiological effluent control program 500 for a nuclear power plant has the following six basic objectives, which are also reflected in 501 10 CFR 50.36a and in site-specific Technical Specifications:

502 503 (1) Ensure that effluent instrumentation has the functional capability to measure and analyze effluent 504 discharges.

505 506 (2) Ensure that effluent treatment systems are used to reduce effluent discharges to ALARA levels.

507 508 (3) Establish instantaneous release-rate limitations on the concentrations of radioactive material.

509 510 (4) Limit the annual and quarterly doses or dose commitment to members of the public in liquid and 511 gaseous effluents to unrestricted areas.

512 513 (5) Measure, evaluate, and report the quantities of radioactivity in gaseous effluents, liquid effluents, RG 1.21, Rev. 3, Page 12

514 and solid radioactive waste shipments.

515 516 (6) Evaluate the dose to members of the public.

517 518 As required by technical specifications, Part 50 and Part 52 licensees must submit the Annual 519 Radioactive Effluent Release Report (ARERR) before May 1 and the AREOR by May 15 of each year 520 (unless a licensing basis exists for a different submittal date for one or both reports). Licensees use these 521 reports to demonstrate compliance with the facilitys technical specifications for the radioactive effluent 522 control program. The reports demonstrate the following:

523 524

• effectiveness of effluent controls and measurement of the environmental impact of radioactive 525 materials 526 527

• compliance with the design objectives and limiting conditions for operation required to meet the 528 ALARA criteria in 10 CFR Part 50, Appendix I 529 530

• relationship between quantities of radioactive material discharged in effluents and resultant 531 radiation dose to individuals 532 533

• compliance with the radiation dose limits to members of the public established by the NRC and 534 the U.S. Environmental Protection Agency (EPA) 535 536

• compliance with the effluent reporting requirements of 10 CFR 50.36a1 537 538 Consideration of International Standards2 539 540 The International Atomic Energy Agency (IAEA) works with member states and other partners to 541 promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety 542 Requirements and Safety Guides for protecting people and the environment from harmful effects of 543 ionizing radiation. These requirements and guides provide a system of Safety Standards Categories that 544 reflect an international perspective on what constitutes a high level of safety. In developing or updating 545 Regulatory Guides the NRC has considered IAEA Safety Requirements, Safety Guides1 and other 546 relevant reports in order to benefit from the international perspectives, pursuant to the Commissions 547 International Policy Statement (Ref. 41) and NRC Management Directive and Handbook 6.6 (Ref. 42).

548 549 The following IAEA Safety Standards Series are consistent with the basic safety principles considered in 550 developing this Regulatory Guide:

551 552

• IAEA General Safety Guide (GSG)-8, Radiation Protection of the Public and the Environment, 553 issued 2018 (Ref. 43) 554 1 See Section C.9 of this regulatory guide for information regarding use of the ARERR or its format to also meet ISFSI effluent reporting requirements in 10 CFR 72.44(d) for specific licenses or imposed by certificate of compliance conditions for general licenses.

2 IAEA Safety Requirements and Guides may be found at https://www.iaea.org/ or by writing the International Atomic Energy Agency, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria; telephone (+431) 2600-0; fax (+431) 2600-7; or e-mail Official.Mail@IAEA.Org. It should be noted that some of the international recommendations do not correspond to the NRC requirements which take precedence over the international guidance.

RG 1.21, Rev. 3, Page 13

555

• IAEA Specific Safety Guide NS-G-3.2, Dispersion of Radioactive Material in Air and Water 556 and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants, 557 issued 2002 (Ref. 44) 558 559

• IAEA GSG-9, Regulatory Control of Radioactive Discharges to the Environment, issued 2018 560 (Ref. 45) 561 562

• IAEA GSG RS-G-1.8, Environmental and Source Monitoring for Purposes of Radiation 563 Protection, issued 2005 (Ref. 46) 564 565

• IAEA Nuclear Energy Series NP-T-3.16, Accident Monitoring Systems for Nuclear Power 566 Plants, issued 2015 (Ref. 47) 567 568

• IAEA-TECDOC-482, Prevention and Mitigation of Groundwater Contamination from 569 Radioactive Releases, Vienna, Austria, issued 1988 (Ref. 48) 570 571

• IAEA Safety Guide No. WS-G-3.1, Remediation Process for Areas Affected by Past Activities 572 and Accidents, Vienna, Austria, issued 2007 (Ref. 49) 573 574

• IAEA, Management of Waste Containing Tritium and Carbon-14, Technical Report Series 575 Number 421, Vienna, Austria, issued 2004 (Ref. 50) 576 RG 1.21, Rev. 3, Page 14

577 C. STAFF REGULATORY GUIDANCE 578 579 1. Effluent Monitoring 580 581 1.1 Effluent Monitoring Programs 582 583 Monitoring programs shall be established to identify and quantify principal radionuclides in 584 effluents in accordance with 10 CFR 50.36a. NUREG-1301 (for PWRs) and NUREG-1302 (for BWRs) 585 provide guidance on acceptable methods of generic controls and surveillance requirements, including 586 frequency, duration, and methods of measurement. These NUREGs provide acceptable LLDs, guidance 587 on batch releases and continuous releases, sampling frequencies, analysis frequencies and timelines, and 588 composite sample guidance. Site-specific radiological effluent control programs that differ from the 589 generic NUREG-1301 and NUREG-1302 guidance should be based on a documented evaluation or 590 justification for such deviations as part of an ODCM authorized change, or, if submitted and approved as 591 part of the original ODCM, in accordance with GL 89-01.

592 593 1.2 Release Points for Effluent Monitoring 594 595 The ODCM (or equivalent), as required by technical specifications, should identify the facilitys 596 significant release points (see definition in the glossary) used to quantify liquid and gaseous effluents 597 discharged to the unrestricted area. For those release points containing contributions from two or more 598 inputs (or systems), it is preferable to monitor each major input (or system) individually to avoid dilution 599 effects, which may impede or prevent radionuclide identification. NUREG-1301 and NUREG-1302 600 contain detailed guidance for the content and format of a licensees ODCM. For purposes of effluent and 601 direct radiation monitoring, the ODCM should list and describe the following:

602 603 1. significant release points (see definition in Section 1.3 and in the glossary), which include stacks, 604 vents, and liquid radioactive waste discharge points, among others; 605 606 2. less-significant release points (see definition in Section 1.4 and in the glossary) that are not 607 normally classified as one of the significant release points but could become a significant release 608 point based on expected operational occurrences (e.g., primary to secondary leakage for PWRs or 609 failed fuel)3; 610 611 3. the site environs map, which should show each of the following:

612 613 a. significant release points, 614 615 b. boundaries of the restricted area and the controlled area4 (in accordance with 616 10 CFR Part 20 definitions),

617 3 This list does not need to be exhaustive or all-inclusive but should demonstrate that the licensee has reasonably anticipated expected operational occurrences and their effects on radioactive discharges. Examples may include main steam line safety valves, steam-driven feedwater pumps, turbine building sumps, containment ice condensers, leachate seepage from unlined ponds, or evaporative releases from ponds in the restricted or controlled areas.

4 For ODCMs that also address Part 72 monitoring requirements, the boundaries of the Part 72 controlled area, as defined in 10 CFR 72.3 and meeting the minimum size requirements of 72.106 should be also be shown.

RG 1.21, Rev. 3, Page 15

618 c. boundary of the unrestricted area5 for liquid effluents (e.g., at the end of the pipe or 619 entrance to a public waterway), and 620 621 d. boundary of the unrestricted area for gaseous effluents (e.g., the site boundary).

622 623 4. dose calculation methodologies for exposure pathways and routes of exposure that are identified 624 in RG 1.109, if applicable; and 625 626 5. dose calculation methodologies for direct radiation if necessary (e.g., when assessing direct 627 radiation from the facility)6.

628 629 1.3 Monitoring a Significant Release Point 630 631 A significant release point is any location from which radioactive material is released that 632 contributes greater than 1 percent of the activity discharged from all the release points for a particular 633 type of effluent considered. RG 1.109 lists the three types of effluent as (1) liquid effluents, (2) noble 634 gases released to the atmosphere, and (3) all other radionuclides discharged to the atmosphere.

635 636 The ODCM should list significant release points. Significant release points should be monitored 637 in accordance with the ODCM. If a new significant release point is identified and is not listed in the 638 ODCM, licensees should (1) establish an appropriate sampling interval (e.g., in site-specific procedures) 639 and (2) update the ODCM within a reasonable timeframe (e.g., annually). Releases from a significant 640 release point should be assessed based on an appropriate combination of actual sample analysis results, 641 radiation monitor responses, flow rate indications, tank level indications, and system pressure indications 642 as necessary to ensure that the amount of radioactive material released, and the corresponding doses, are 643 not substantially underestimated (see 10 CFR Part 50, Appendix I, Section III, Implementation). If 644 activity is detected when monitoring a significant release point, the radionuclides detected should be 645 reported in the effluent totals (including those with half-lives less than 8 days) in the ARERR (i.e., in 646 Table A-1 or Table A-2), provided that the amount discharged is significant to the three-digit exponential 647 format required for the ARERR.

648 649 1.4 Monitoring a Less-Significant Release Point 650 651 NUREG-1301 and NUREG-1302 provide tables designating sampling and analysis frequencies 652 for release points. Historically, these tables, together with the guidance from RG 1.21, Revision 1, issued 653 June 1974 (Ref. 51) or RG 1.21, Revision 2, issued June 2009 (Ref. 52) provide sampling and analysis 654 frequencies. Licensees may continue to use the guidance from NUREG-1301 or NUREG-1302 and/or 655 Revision 1 or Revision 2 of RG 1.21 in accordance with their ODCMs. This method of assigning sample 656 frequencies is simple to implement but, in certain cases, may entail an inappropriately large number of 657 samples for less-significant release points with noor extremely lowimpact on the parameters reported 658 in the ARERR. As a result, for less-significant release points, licensees may evaluate and assign more 659 appropriate sampling frequencies. If a licensee wishes to deviate from the NUREG-1301 and NUREG-660 1302 sampling frequencies, the licensees evaluation must show that the changes (i.e., deviations from 661 NUREG-1301 and NUREG-1302) maintain the levels of radioactive effluent control as stated in the 662 technical specifications required by 10 CFR 20.1302; 40 CFR Part 190; 10 CFR 50.36a; and 5 The boundaries of the unrestricted areas may be defined separately for liquid effluents, gaseous effluents, and if appropriate, for other radiological controls such as direct radiation.

6 The methodology should include background subtraction, and if appropriate, extrapolation of radiation measurements to points of interest (e.g., to the individual members of the public likely to receive the highest dose).

RG 1.21, Rev. 3, Page 16

663 10 CFR Part 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or 664 setpoint calculations, and should be maintained in site documentation. Regardless of the surveillance 665 frequencies, if activity is detected when monitoring a less-significant release point, the licensee must (in 666 accordance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section III.A.1) report the cumulative 667 activity in the effluent totals (i.e., in Table A-1 or Table A-2) in the ARERR (provided that the amount 668 discharged is significant to the three-digit exponential format required for the ARERR).

669 670 Site documentation should identify less-significant release points, to the extent reasonable, but it 671 is not necessary to list all possible release points in site documentation. Releases from a less-significant 672 release point may be assessed (see Section 5.1) to the extent reasonable using assumptions and bounding 673 calculations (in lieu of, or in addition to, sampling and analysis). When plant conditions change and such 674 changes may reasonably affect the status of a less-significant release point (e.g., significant change in 675 primary-to-secondary leakage in PWRs or substantial cross contamination between systems), the licensee 676 should sample and analyze the affected less-significant release points. These sample results should be 677 evaluated to (1) confirm the continued validity of the bounding calculations (if used) with regard to 678 effluent accountability and (2) determine the impact (if any) on effluent accountability. The guidance in 679 this RG on monitoring less-significant release points for purposes of accountability (through the ARERR) 680 does not replace, supersede, or otherwise modify any responsibility for monitoring systems normally not 681 contaminated, as outlined in NRC Inspection and Enforcement Bulletin 80-10, Contamination of 682 Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity 683 to Environment, issued May 1980 (Ref. 53). A thoroughly designed and documented evaluation of a 684 less-significant release point could also assist in the evaluation and characterization of abnormal releases 685 and abnormal discharges (see Section 1.11 below).

686 687 1.5 Monitoring Leaks and Spills 688 689 An area where an unplanned release occurred in the onsite environs (e.g., a leak or spill) should 690 be identified as an impacted area, as defined in 10 CFR 50.2, Definitions, for decommissioning 691 purposes, and in accordance with NUREG-1757. A leak or spill should be assessed to obtain the 692 necessary information for the ARERR, as specified in Section 8.5.1 of this RG (see glossary).

693 694 Leaks or spills to the ground and/or subsurface will be diluted on contact with soil and water in 695 the environment; therefore, samples of the undiluted liquid (from the source of the leak or spill) and 696 samples of the affected soil (or surface water or subsurface groundwater) should be analyzed as soon as 697 practical. In some instances, sampling, particularly soil sampling, may not be practical if the leak 698 occurred in inaccessible areas or if there are extenuating considerations. In this respect, groundwater 699 monitoring may be used as a surrogate for soil sampling. If sampling is not practical, the 700 10 CFR 50.75(g) records should describe why sampling was not conducted (e.g., the area was 701 inaccessible or there were safety considerations). The licensee should ensure that the location and 702 estimated volume of the leak or spill are recorded to identify the extent of the impacted area and predicted 703 size or extent of the contaminant plume, both horizontally and vertically. If a spill is promptly and fully 704 remediated (e.g., within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) and if subsequent surveys of the remediated area indicate no detectable 705 residual radioactivity remaining in the soil or groundwater (see paragraph below), for purposes of 706 reporting discharges in the ARERR, there was no liquid discharge to the unrestricted area, and the spill 707 need not be reported in the ARERR. However, in accordance with 10 CFR 50.75(g), the 708 decommissioning file should be updated to include a description of the leak or spill event. Licensees 709 should review the decommissioning files before generating the ARERR to ensure that the ARERR 710 includes the necessary information on leaks and spills.

711 712 When evaluating areas that have been remediated, the licensee should survey for residual 713 radioactivity. There may be times when the licensee wants to verify that an area contains no residual RG 1.21, Rev. 3, Page 17

714 radioactivity. There is existing regulatory guidance and information on analytical detection capabilities.

715 Licensees should ensure that surveys are appropriate and reasonable, in accordance with 10 CFR 20.1501.

716 Licensees should generally ensure that surveys are conducted using the appropriate sensitivity 717 levels; e.g., refer to the environmental LLDs in NUREG-1301 and NUREG-1302, Table 4.12-1, 718 Detection Capabilities for Environmental Sample Analysis, or LLDs determined by using the 719 methodology outlined in NUREG-1576. Additionally, licensees should apply plant-process-system 720 knowledge when evaluating leaks and spills.

721 722 This RG provides guidance on information that licensees should provide in the ARERR. In that 723 context, when leaks and spills of radioactive material are identified, prompt response and timely actions 724 should be taken to the extent reasonable to (1) evaluate onsite radiological conditions and (2) ensure 725 proper reporting of materials discharged off site. To realize these two goals, it may be necessary to 726 isolate the leak or spill at the source, prevent the spread of the leak or spill, and remediate the affected 727 area (if the licensee deems remediation to be reasonable and necessary).

728 729 For leaks and spills involving the discharge of radioactive material to an unrestricted area, 730 licensees should follow RG 4.25 or equivalent methods to assess the amount of material discharged to the 731 unrestricted area. The potential dose to members of the public from the leak or spill should be evaluated 732 using realistic or bounding exposure scenarios. Attachment 6 to SECY-03-0069, Results of the License 733 Termination Rule Analysis, dated May 23, 2003 (Ref. 54), provides more information on the use of 734 realistic scenarios.

735 736 For leaks and spills, licensees should perform surveys that are reasonable to evaluate the potential 737 radiological hazard (as described in 10 CFR 20.1501). As a result, for leaks and spills, licensees may 738 choose to use bounding assessments to estimate the potential hazard. For example, if a leak occurs on site 739 and radioactive material is released at or below the ground surface, the licensee may choose to assess the 740 potential hazard by estimating a conservatively large (e.g., bounding) volume of water as part of an 741 assumed exposure pathway analysis (e.g., drinking water). Such assumptions would allow the licensee to 742 assess the potential hazard to a hypothetical individual member of the public. A hazard assessment of this 743 sort would be appropriate for inclusion in the supplemental information section of the ARERR. If there is 744 no real exposure pathway to a member of the public, the licensee should indicate that the hazard 745 assessment is a bounding estimate of the dose to a hypothetical individual member of the public, and no 746 real individual member of the public received an actual exposure.

747 748 If licensees choose to notify local authorities of spills or leaks (e.g., because of local ordinances 749 or local and State government agreements), the licensee should review the reporting requirements of 750 10 CFR 50.72(b)(xi) and information in NUREG-1022, Event Reporting Guidelines: 10 CFR 50.72 and 751 50.73, issued October 2000 (Ref. 55), for applicability. In such situations, licensees should ensure 752 effective communication, using NUREG/BR-0308, Effective Risk Communication, issued June 2004 753 (Ref. 56), especially when ensuring that the risk is described in the appropriate context. In general, 754 licensees should notify the NRC when significant public concern is raised, in accordance with 755 10 CFR 50.72(b)(xi).

756 757 Although the licensee may choose to use its problem identification and resolution program 758 (corrective action program) to document the evaluation of the spill or leak, appropriate documentation 759 should be placed in, or cross referenced to, the decommissioning files, as required by 10 CFR 50.75(g).

760 761 Although prompt remediation is not a requirement (Ref. 57), remediation should be evaluated and 762 implemented, as appropriate, based on licensee evaluations and risk-informed decisionmaking. The 763 Electric Power Research Institute (EPRI) Report 1021104 Groundwater and Soil Remediation 764 Guidelines for Nuclear Power Plants, proprietary report issued December 2010 (Ref. 58) and EPRI RG 1.21, Rev. 3, Page 18

765 Report 1023464, Groundwater and Soil Remediation Guidelines for Nuclear Power Plants, (Public 766 Edition) Final Report, July 2011 (Ref. 59) may be useful in performing remediation evaluations.

767 768 Evaluation factors should include (1) the location and accessibility, (2) the concentrations of 769 radionuclides and extent of the residual radioactivity, (3) the efficacy of monitored natural attenuation, 770 (4) the volume of the release, (5) the mobility of the radionuclides, (6) the depth of the water table, and 771 (7) whether significant residual radioactivity (see glossary) is expected at the time of 772 decommissioning. Since the contaminants, concentrations, and extent of contamination are expected to 773 vary over time or plant life (either increase based on anticipated future leaks and spills or decrease based 774 on remediation or monitored natural attenuation), no one set of numerical values defines significant 775 residual radioactivity. However, licensees may make remediation decisions based on their expectations 776 of their ability to meet the decommissioning criteria of 10 CFR 20.1402 at the anticipated time of 777 decommissioning.

778 779 Information that may be useful in this risk-informed decision making includes (1) NUREG-1757, 780 Volume 1, Appendix H, EPA/NRC Memorandum of Understanding, (2) NUREG-1757, Volume 2, 781 Table H.1, Acceptable License Termination Screening Values of Common Radionuclides for 782 Building-Surface Contamination, and (3) the authorized derived concentration guideline levels for 783 decommissioned nuclear power plants. For a more detailed analysis, licensees may use the computer 784 codes described in NUREG/CR-6676, Probabilistic Dose Analysis Parameter Distributions Developed 785 for RESRAD and RESRAD-BUILD Codes, issued July 2000 (Ref. 60); NUREG/CR-6692, 786 Probabilistic Modules for the RESRAD and RESRAD-BUILD Computer Codes, issued November 787 2000 (Ref. 61); NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 788 3.0 Computer Codes, issued December 2000 (Ref. 62); and NUREG/CR-7267, Default Parameter 789 Values and Distributions in RESRAD-ONSITE V7.2, RESRAD-BUILD V3.5 and RESRAD-OFFSITE 790 4.0 (Ref. 63).

791 792 1.6 Monitoring Continuous Releases 793 794 For continuous releases, gross radioactivity measurements are often the only practical means of 795 continuous monitoring. These gross radioactivity measurements are typically used to actuate alarms and 796 terminate (trip) effluent releases; by themselves, such measurements are generally not acceptable for 797 demonstrating compliance with effluent discharge limits.

798 799 The use of continuously indicating radiation monitoring system results may be combined with 800 sample analyses to more fully characterize and quantify a discharge. This technique may have particular 801 applicability when (1) a short-term, rapid upscale indication of a process radiation monitor occurs during 802 a release or (2) when there is a desire to verify whether a preliminary grab sample is representative. In 803 these instances, the licensee should ensure that the radiation monitor responses (i.e., the radiation monitor 804 efficiencies) for various radionuclides are well characterized.

805 806 Grab samples should be collected at scheduled frequencies in accordance with the ODCM (see 807 NUREG-1301 and NUREG-1302 or as approved in GL 89-01 submittals) to quantify specific 808 radionuclide concentrations and release rates. The frequency of sample collection and radionuclide 809 analyses should be based on the degree of variance in (1) the magnitude of the discharge and (2) the 810 relative radionuclide composition from an established norm. If the magnitude of the discharge and the 811 relative nuclide composition of a continuous release vary significantly over the course of the discharge 812 period, a combination of grab samples and continuous monitor readings can assist in accurately 813 estimating the discharge. Continuous monitoring data (e.g., chart recorder data), as well as grab sample 814 data, should be reviewed periodically and used to identify this variance from the established norm.

815 Periodic evaluations should be made between gross radioactivity measurements and grab sample analyses RG 1.21, Rev. 3, Page 19

816 of specific radionuclides. These evaluations should be used to verify (or modify) the conversion factors 817 that correlate radiation monitor readings and concentrations of radionuclides in effluents.

818 819 NUREG-1301 and NUREG-1302 provide guidance on the Radiological Environmental 820 Monitoring Program. Table 3.12-1 therein provides guidance on implementing the environmental 821 monitoring program, including an I-131 sampling and analysis on each composite of drinking water.

822 823 If a drinking water exposure pathway exists, a prospective dose evaluation should be performed 824 based on I-131 in effluent discharges to determine the maximum likely annual I-131 thyroid dose to a 825 person in any age group. The purpose of the prospective dose evaluation is to determine the 826 environmental sampling and analysis requirements. Note: Freshwater fish ingestion is not included in 827 the prospective dose evaluation of I-131 from the drinking water route of exposure.

828 829 If the likely dose from I-131 is greater than 1 mrem per year, a composite drinking water sample 830 should be collected over a 2-week period and an I-131 analysis performed with an LLD of 1 pCi/liter. If 831 the likely dose from I-131 is less than or equal to 1 mrem per year, a monthly composite sample should be 832 collected and an I-131 analysis performed with an LLD of 15 pCi/liter.

833 834 In addition, Standard Technical Specifications require determination of the projected dose 835 contributions from radioactive effluents at least every 31 days, and determination of the cumulative dose 836 contributions for the current calendar quarter and current calendar year.

837 838 1.7 Monitoring Batch Releases 839 840 For batch releases, measurements should be performed to identify principal radionuclides before 841 a release. If an analysis of specific hard-to-detect radionuclides (such as strontium-89/90, Ni-63 and 842 iron-55 in liquid releases) cannot be done before the batch release (see NUREG-1301 and NUREG-1302),

843 the licensee should have collected representative samples for the purpose of subsequent composite 844 analysis. The composite samples should be analyzed at the scheduled frequencies specified in 845 NUREG-1301 and NUREG-1302 or at the revised frequencies specified by the licensee (with 846 documented justification in accordance with ODCM change process specified in the technical 847 specifications) (see Sections 1.3 and 1.4 of this RG).

848 849 Continuously indicating radiation monitoring system results may be combined with sample 850 analyses to more fully characterize and quantify a discharge. This technique may have particular 851 applicability when (1) a short-term, rapid upscale indication of a process radiation monitor occurs during 852 a discharge or (2) when there is a desire to verify whether a preliminary grab sample is representative. In 853 these instances, the licensee should ensure that radiation monitor responses (i.e., the radiation monitor 854 efficiencies) for various radionuclides are well characterized.

855 856 1.8 Principal Radionuclides for Effluent Monitoring 857 858 This RG introduces the term principal radionuclide in a risk informed context. A licensee may 859 evaluate the list of principal radionuclides for use at a particular site. The principal radionuclides maybe 860 determined based on their relative contribution to either (1) the public dose compared to the 861 10 CFR Part 50, Appendix I, design objective doses, or (2) the amount of activity discharged compared to 862 other site radionuclides in the type of effluent being considered. Under this concept, radionuclides that 863 have either a significant activity or a significant dose contribution should be monitored in accordance 864 with a predetermined and appropriate analytical sensitivity level (LLD) outlined in a licensees ODCM.

865 This implementation of principal radionuclides ensures that the ARERR appropriately includes both the 866 (1) radionuclides that are present in relatively large amounts but that contribute very little to dose and RG 1.21, Rev. 3, Page 20

867 (2) radionuclides that are present in very small amounts but that have a relatively high contribution to 868 dose.

869 870 If a risk-informed approach is used, principal radionuclides should be determined based on an 871 evaluation over a time period that includes a refueling outage (e.g., one fuel cycle). A periodic 872 reevaluation should be performed to determine whether the radionuclide mix has changed and to identify 873 new principal radionuclides.

874 875 If a risk-informed approach is applied to the determination of principal radionuclides,7 the ODCM 876 becomes the controlling document and specifies the list of principal radionuclides. If adopting this 877 method, the licensee should update the ODCM with the list of principal radionuclides within 1 year of 878 their identification. Licensees are allowed to revise the ODCM in accordance with the ODCM change 879 process, as described in the plants technical specifications (which includes documented evaluations of 880 such changes).

881 882 If adopting a risk-informed approach, a radionuclide is considered a principal radionuclide if it 883 contributes either (1) greater than 1 percent of the 10 CFR Part 50, Appendix I, design objective dose for 884 all radionuclides in the type of effluent being considered or (2) greater than 1 percent of the activity of all 885 radionuclides in the type of effluent being considered. RG 1.109 lists the three types of effluent as 886 (1) liquid effluents, (2) noble gases released to the atmosphere, and (3) all other radionuclides released to 887 the atmosphere. In this context, the term principal radionuclide has special significance for the required 888 sensitivity levels (e.g., LLDs) for an analysis. The LLDs specified in NUREG-1301 and NUREG-1302 889 may be used, or LLDs may be determined based on the other methodologies (e.g., as outlined in 890 NUREG-1576). Once principal radionuclides are identified, they should be monitored in accordance with 891 the sensitivity levels (e.g., LLDs) listed in the ODCM.

892 893 During analysis of samples, licensees should apply the appropriate analytical sensitivities to 894 ensure adequate surveys are conducted. NUREG-1301 and NUREG-1302 provide a list of principal 895 gamma emitters for operating reactors for which an LLD control applies. Historically, this list and 896 guidance from Revision 1 or Revision 2 provided the appropriate sensitivity levels for an analysis.

897 Licensees may continue to use this historical guidance, which essentially classifies all radionuclides as 898 principal radionuclides, and apply the analytical sensitivity levels (e.g., LLDs) directly from 899 NUREG-1301 and NUREG-1302 and Revision 1 or 2 of RG 1.21. This method is simple to implement 900 but, in certain cases, may entail inappropriately long count times or may involve alternate (or 901 unnecessary) methods of analysis for low-activity radionuclides with noor extremely lowdose 902 significance.

903 904 Although the LLD list from NUREG-1301 and NUREG-1302 may be used to determine principal 905 radionuclides, in reality, the principal radionuclides at a site will depend on site-specific factors, such as 906 (1) the operating status of the facility (e.g., operating or in decommissioning), (2) the amount of failed 907 fuel, (3) the extent of system leakage, (4) the sophistication of radioactive waste processing equipment, 908 and (5) the level of expertise in operating radioactive waste processing systems. Since the principal 909 radionuclides will vary from site to site, licensees that wish to deviate from the historical method of 910 determining principal radionuclides (as described above) may adopt a risk-informed approach to identify 911 principal radionuclides (and the associated sensitivity levels) at a site.

912 913 For radionuclides that are not identified as principal radionuclides, licensees may use their 914 discretion with the sensitivity of analysis, provided the licensees determine that the changes maintain the 7 With respect to principal radionuclides, dose is the measure of risk, whereas activity is not. For example, a relatively large amount of tritium released into a large body of water has little dose significance.

RG 1.21, Rev. 3, Page 21

915 levels of radioactive effluent controls required by the regulations in 10 CFR 20.1302; 40 CFR Part 190; 916 10 CFR 50.36a; and 10 CFR Part 50, Appendix I, and do not adversely impact the accuracy or reliability 917 of effluent, dose, or setpoint calculations. If licensees change their analytical sensitivities from those in 918 their ODCM or equivalent, they must document the basis for the deviations. For example, DQOs and 919 other concepts from RG 4.15 may be useful for determining risk-informed sensitivity levels for an 920 analytical method.

921 922 The risk-informed concept of principal radionuclides does not reduce the requirement for 923 reporting radionuclides detected in effluents. In addition to principal radionuclides, other radionuclides 924 detected during routine monitoring of release points should be reported in the radioactive effluent release 925 report and included in dose assessments to members of the public, consistent with site-specific technical 926 specifications.

927 928 1.9 Carbon-14 929 930 Carbon (C)-14 is a naturally occurring isotope of carbon. Nuclear weapons testing in the 1950s 931 and 1960s significantly increased the amount of C-14 in the atmosphere. Commercial nuclear reactors 932 also produce C-14 but in much lower amounts than those produced naturally or from weapons testing.

933 IAEA Report Number 421 provides relevant information on C-14 releases. The C-14 releases in PWRs 934 occur primarily as a mix of organic carbon and carbon dioxide released from the waste gas system. In 935 BWRs, C-14 releases occur mainly as carbon dioxide in gaseous waste.

936 937 Regulations in 10 CFR 50.36a require that operating procedures be developed for the control of 938 effluents and that quantities of principal radionuclides be reported. The radioactive effluents from 939 commercial nuclear power plants overtime has decreased to the point that C-14 is likely to have become a 940 principal radionuclide (as defined in this document) in gaseous effluents. Therefore, licensees must 941 evaluate whether C-14 is a principal radionuclide for gaseous releases from their facility. Because the 942 dose contribution of C-14 from liquid radioactive waste is much less than that contributed by gaseous 943 radioactive waste, an evaluation of C-14 in liquid radioactive waste is not required.

944 945 The quantity of C-14 discharged can be estimated by use of a normalized C-14 source term and 946 scaling factors based on power generation or estimated by use of the NUREG-0016 (GALE-BWR) and 947 NUREG (GALE-PWR) computer codes. The National Council on Radiation Protection and 948 Measurements Report No. 81, Carbon-14 in the Environment, (Ref. 64) also provides information about 949 the magnitude of C-14 in typical effluents from commercial nuclear power plants. These documents 950 estimate that nominal annual releases of C-14 in gaseous effluents are approximately from 5 to 7.3 curies 951 from PWRs and from 8 to 9.5 curies from BWRs.

952 953 The quantity of C-14 generated in BWR and PWR cores can also be estimated by a calculational 954 method provided by the EPRI Report No. 1021106, Estimation of Carbon-14 in Nuclear Power Plant 955 Gaseous Effluents, issued December 2010 (Ref. 65) and EPRI Report No. 1024827 "Carbon-14 Dose 956 Calculation Methods at Nuclear Power Plants," issued April 2012, (Ref. 66). If estimating C-14 based on 957 scaling factors and fission rates, a precise and detailed evaluation of C-14 is not necessary. It is not 958 necessary to calculate uncertainties for C-14 or to include C-14 uncertainty in any subsequent calculation 959 of overall uncertainty.

960 961 Since the NRC published RG 1.21, Revision 1, in 1974, the analytical methods for determining 962 C-14 have improved. Because the production of C-14 is expected to be relatively constant at a particular 963 site, if sampling is performed for C-14 (instead of estimating C-14 discharges based on calculations), the 964 sampling frequency may be adjusted to that interval that allows adequate measurement and reporting of 965 effluents.

RG 1.21, Rev. 3, Page 22

966 967 1.10 Return/Reuse of Previously Discharged Radioactive Effluents 968 969 Radioactive material properly released in gaseous or liquid effluents to the unrestricted area 970 (excluding solid materials or soil) is not considered licensed material when returned to the facility as long 971 as the concentration of radioactive material does not exceed 10 CFR Part 30, Rules of General 972 Applicability to Domestic Licensing of Byproduct Material, exempt concentration limits (otherwise a 973 general or specific license is required). The water containing radioactive material returned from the 974 environment can be used by the licensee and returned to the unrestricted area without being considered a 975 new radioactive material effluent release. The basis for this determination is that the licensee has already 976 accounted for this radioactive material when the effluent was originally discharged, provided that the 977 subsequent use, possession, or release does not introduce a new significant dose pathway to a member of 978 the public, as explained below.

979 980 Licensees are responsible for evaluating any new significant exposure pathway and the resultant 981 radiological hazards associated with the return of radioactive material to the operating facility and its 982 subsequent discharge to the environment. For purposes of estimating dose during operations or 983 decommissioning, a new significant exposure pathway is any pathway that contributes dose that exceeds 984 10% of the dose criteria in 10 CFR 50 Appendix I, Section II (such that the dose from a new exposure 985 pathway is unlikely to be substantially underestimated). Bounding dose assessments as described in 986 Section 5.1 of this RG may be used in evaluating any new significant exposure pathway. Furthermore, 987 before returning radioactive materials to the environment, licensees must demonstrate that these 988 radioactive materials were previously disposed of in accordance with 10 CFR 20.2001(a)(3), or that the 989 material is naturally occurring background radiation. Radioactive material previously not accounted for as 990 an effluent that is entrained with returned/re-used water must be considered a new effluent disposal per 10 991 CFR 20.2001. See RIS 2008-03 for further details.

992 993 1.11 Abnormal Releases and Abnormal Discharges 994 995 In RG 1.21, Revision 1, the terms release and discharge were synonymous. In RG 1.21, 996 Revision 2 and 3, the term release describes an effluent from the plant (regardless of where the effluent 997 is located), and the term discharge describes an effluent that enters the unrestricted area. Although the 998 term release includes effluents to either (1) the onsite environs or (2) the unrestricted area, this RG 999 generally reserves use of the term release for the release of an effluent from the power plant into the 1000 onsite environs. The onsite environs in this context encompass locations outside of nuclear power plant 1001 systems, structures, and components, as described in the final safety analysis report or ODCM. This is a 1002 change in terminology with respect to the definition of abnormal release in RG 1.21, Revision 1, which 1003 defined abnormal releases to be from the site boundary.

1004 1005 An abnormal release (see glossary) is an unplanned or uncontrolled release of licensed 1006 radioactive material into the onsite environs. Abnormal releases may be categorized as either batch or 1007 continuous, depending on the circumstances. By contrast, an abnormal discharge (see glossary) is an 1008 unplanned or uncontrolled discharge of licensed radioactive material to the unrestricted area. Abnormal 1009 discharges may also be categorized as either batch or continuous, depending on the circumstances. The 1010 distinction between the terms abnormal release and abnormal discharge is important for describing 1011 the staff position for measuring, evaluating, and reporting releases and discharges, especially where leaks 1012 and spills are involved.

1013 1014 That portion of an abnormal release discharged to the unrestricted area is reported as an abnormal 1015 discharge in the year in which the discharge to the unrestricted area occurred. The portion of an abnormal RG 1.21, Rev. 3, Page 23

1016 release that remains onsite is considered residual radioactivity (see 10 CFR Part 20) and is documented in 1017 accordance with 10 CFR 50.75(g).

1018 1019 Low-level radioactive system leakage resulting from minor equipment failures and component 1020 aging (wear and tear) may be expected to occur as an anticipated part of the plant operation. If such 1021 leakage is captured by, or directed to, a system designed to accept and handle radioactive material, 1022 including the subsequent planned and controlled discharge of the radioactive material (e.g., as described 1023 in the final safety analysis report or ODCM), that evolution is not considered an abnormal release.

1024 Normal system leakage captured by effluent ventilation control systems or sumps is not an abnormal 1025 release (provided that, before discharge of the radioactive material, the discharge is planned and 1026 controlled). (See also the definitions of unplanned release and uncontrolled release in the glossary.)

1027 1028 In certain circumstances, some subjectivity may be associated with the definitions of unplanned 1029 release and uncontrolled release. In these situations, additional circumstances should be considered to 1030 determine whether an abnormal release occurred. A well-designed and documented evaluation of a 1031 release point can include an evaluation of the potential for an unplanned or uncontrolled release. The 1032 evaluation can establish bounding criteria that establish a threshold for an abnormal release based on 1033 planning and control. Generally, releases that may reasonably be categorized as both unplanned and 1034 uncontrolled should be considered abnormal releases.

1035 1036 For example, consider an underground pipe that carries radioactive liquid to an outside storage 1037 tank. If this pipe develops a leak, and licensed radioactive material escapes into the surrounding soil, it is 1038 considered an abnormal release if some portion or all of the radioactive material remains onsite. This 1039 type of leak should be reported as an abnormal release in the next ARERR. If the licensee predicts 1040 (e.g., based on site conceptual model and subsequent groundwater monitoring results) that the radioactive 1041 material will enter the unrestricted area in 2 years, the resulting radioactive discharge (that would occur 1042 2 years hence) will be considered an abnormal discharge. Therefore, the resulting radioactive discharge 1043 should be reported along with other data for the affected calendar year in a future ARERR (i.e., in this 1044 example, 3 years later). Both releases and discharges (either routine or abnormal) should be reported on a 1045 calendar-year basis for the year in which the release or discharge occurred.

1046 1047 Consider another example involving a volume of radioactive gas from the containment 1048 atmosphere that escapes the equipment hatch during a refueling outage (especially during the time 1049 interval when the containment purge exhaust fans are off). This would generally not be considered an 1050 abnormal discharge if (1) the duration was preplanned (e.g., for a short duration such as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />),

1051 (2) the containment activity (gas, particulate, tritium, and iodine) was preplanned, known, and very low 1052 (e.g., such that a bounding estimate of the radioactive material discharged indicated there would be no 1053 measurable impact relative to typical discharges), (3) the containment activity was monitored (e.g., by 1054 sampling or radiation monitoring equipment), and (4) an evaluation was completed to identify a 1055 preplanned limiting (or trigger) level of activity that would initiate remedial or mitigating action 1056 (e.g., close the equipment hatch to control gases escaping containment). In this example, the actions 1057 taken (i.e., preplanning and monitoring) before and during the evolution are sufficient to establish control 1058 of this discharge. As a result, this type of evolution should not be categorized as an abnormal discharge.

1059 1060 2. Effluent Sampling 1061 1062 2.1 Representative Sampling 1063 1064 NUREG-1301 and NUREG-1302 provide a typical schedule for radioactive effluent sample 1065 collection and analyses. Some licensees may have modified these sampling schedules (typically 1066 contained in the ODCM) as part of implementing GL 89-01, as approved by the NRC. Additional RG 1.21, Rev. 3, Page 24

1067 samples should be obtained as needed to characterize abnormal releases, abnormal discharges, or other 1068 significant operational evolutions. Samples should be representative of the overall effluent in the bulk 1069 stream, collection tank, or container. Licensees should ensure that representative samples were obtained 1070 from well-mixed streams or volumes of effluent at sampling points, using proper equipment and sampling 1071 procedures.

1072 1073 2.2 Sampling Liquid Radioactive Waste 1074 1075 Before sampling, large volumes of liquid waste should be mixed to ensure that sediments or 1076 particulate solids are distributed uniformly in the waste mixture. For example, a large tank may be mixed 1077 using a sparger system or recirculated three or more volumes to ensure that a representative sample can be 1078 obtained, as recommended by American Society for Testing and Materials (ASTM) D 3370 - 18, 1079 Standard Practices for Sampling Water from Flowing Process Streams (Ref. 67). If tank-mixing 1080 practices deviate from industry standards (i.e., those for recirculation or otherwise), the licensee should 1081 provide a technical evaluation or other justification. Sample points should be located where there is a 1082 minimum of disturbance of flow caused by fittings and other physical characteristics of the equipment 1083 and components. Sample nozzles should be inserted into the flow or liquid volume to ensure sampling of 1084 the bulk volume of pipes and tanks. Sample lines should be flushed for a sufficient period of time before 1085 sample extraction to remove sediment deposits and air and gas pockets. Generally, three sample line 1086 volumes should be purged as recommended by ASTM D3370 - 18, before withdrawing a sample, unless a 1087 technical evaluation or other justification is provided. A series of samples should be taken periodically 1088 during the interval of discharge to determine whether any differences exist as a function of time and to 1089 ensure that individual samples are indeed representative of the effluent mixture. In some instances, this 1090 may be accomplished by collecting one or more samples (either by grab or composite sampler) during 1091 the discharge and comparing with one or more samples taken before the discharge. If a series of samples 1092 is collected, these samples can be used to assess the amount of measurement uncertainty in obtaining 1093 representative samples.

1094 1095 2.3 Sampling Gaseous Radioactive Waste 1096 1097 Although all licensees may not be committed to RG 4.15, ANSI N42.18-2004, Specification and 1098 Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents 1099 (Ref. 68), ANSI N42.54-2018, Instrumentation and Systems for Monitoring Radioactivity (Ref. 69),

1100 and ANSI/Health Physics Society (HPS) N13.1-2011, Sampling and Monitoring Releases of Airborne 1101 Radioactive Substances from the Stacks and Ducts of Nuclear Facilities (Ref. 70), these documents 1102 provide general principles for designing and conducting monitoring programs for airborne effluents. The 1103 cited references also contain recommendations for obtaining valid samples of airborne radioactive 1104 material in effluents and the guidelines for sampling from ducts and stacks. Licensees should use the 1105 appropriate licensing documents to evaluate the validity of representative samples (e.g., evaluate the 1106 potential for inaccurate sampling of gaseous effluents that may bypass a particulate filter and collect on an 1107 iodine collection cartridge) and to identify any inaccurate sample analyses configurations or counting 1108 geometries.

1109 1110 2.4 Sampling Bias 1111 1112 Sampling and storage techniques that could bias quantitative results for effluent measurements 1113 should be evaluated and corrections applied as necessary. These biases include inaccurate measurement 1114 of sample volumes resulting from pressure drops in long sample lines and loss of particulates or iodine in 1115 sample lines resulting from deposition or plate-out. Samplers for gaseous waste should be evaluated for 1116 particulate deposition using ANSI/HPS N13.1-1999 or equivalent.

1117 RG 1.21, Rev. 3, Page 25

1118 2.5 Composite Sampling 1119 1120 Composite samples should be representative of the average quantities and concentrations of 1121 radioactive materials discharged in liquid and gaseous effluents. Composite samples should be collected 1122 in proportion to the effluent flow rate or in proportion to the volume of each batch of effluent discharges.

1123 1124 2.6 Sample Preparation and Preservation 1125 1126 Sample preparation and storage methods should minimize the potential for loss of radioactive 1127 material (i.e., deposition of analyte on walls of the sample container or volatilization of analyte).

1128 Composite sample storage time should be as short as practical to preclude deposition on the storage 1129 container, or sample stabilization should be considered. Before quantitative radionuclide analyses for 1130 liquid effluent composites, licensees should ensure that samples are mixed thoroughly so that the sample 1131 is representative of the material discharged.

1132 1133 Procedures for handling, packaging, and storing samples should ensure that losses of radioactive 1134 materials or other factors causing sample deterioration do not invalidate the analysis. For example, filters 1135 should be stored carefully to prevent loss of radioactive material from the filter paper.

1136 1137 2.7 Short-Lived Radionuclides and Decay Corrections 1138 1139 In the analysis of short-lived radionuclides (e.g., short-lived noble gases), measurements should 1140 generally be made as soon as practical after collection to minimize loss by radioactive decay. In other 1141 cases, when needed to improve the detection of the longer-lived radionuclides, time should be allowed for 1142 the decay of short-lived, interfering radionuclides.

1143 1144 Some special considerations may be applicable when measuring short-lived radionuclides. In 1145 general, sample collection (or analysis frequencies) should take into account the half-lives of the 1146 radionuclides being measured. This may have special applicability for continuous samples or composite 1147 samples. It is generally best to select a compositing interval (and analysis frequency) appropriate for the 1148 effluent (radionuclide) being analyzed. In cases where the compositing interval is selected appropriately, 1149 analytical bias is minimized. One way to avoid analytical bias is to decrease the composite sampling 1150 interval (and analysis frequency).

1151 1152 To minimize bias in measurements, it may be necessary to decay correct analysis results for 1153 short-lived radionuclides. Licensees should be cognizant of those situations in which analytical bias may 1154 be introduced when analyzing short-lived radionuclides and should select appropriate methods to 1155 minimize such bias.

1156 1157 3. Effluent Dispersion (Meteorology and Hydrology) 1158 1159 3.1 Meteorological Data 1160 1161 Gaseous effluents discharged into the atmosphere are transported and diffused (or, in combination 1162 dispersed and, therefore diluted) as a function of (1) the atmospheric conditions in the local environment 1163 (including ambient meteorology and structural wake effects), (2) the topography of the region, and (3) the 1164 release characteristics of the effluents. In developing and implementing a monitoring program designed 1165 to collect site-specific meteorological data, licensees should, conform to the guidance consistent with 1166 their facilitys current licensing basis but should also consider adopting the guidance in the current 1167 version of RG 1.23. The meteorological data do not need to be reported in the ARERR, but the data 1168 should be summarized and maintained as documentation (records). Licensees should prepare and RG 1.21, Rev. 3, Page 26

1169 maintain an annual meteorological summary report that provides the joint frequency distributions of wind 1170 direction and wind speed by atmospheric stability class (see RG 1.23, or, if applicable, Safety Guide 23, 1171 Onsite Meteorological Programs, dated February 17, 1972 (Ref. 71)) on site for the life of the plant. In 1172 addition, the licensee should record hourly meteorological data (or shorter-term averages compatible with 1173 the appropriate dispersion models) and make the data available if needed for assessing abnormal gaseous 1174 releases.

1175 1176 3.2 Atmospheric Dispersion (Transport and Diffusion) 1177 1178 Site-specific meteorological data collected should be validated and used to generate gaseous 1179 effluent dispersion factors (/Q) and deposition factors (D/Q), in accordance with RG 1.111. The use of 1180 long-term annual-average meteorological conditions (based on 5 or more years of data) to determine /Q 1181 and D/Q is appropriate for continuous releases and for establishing instantaneous release rate set points.

1182 This practice may also be acceptable for calculating doses from intermittent releases if the releases occur 1183 randomly and with sufficient frequency to justify the use of annual-average meteorological conditions 1184 (see RG 1.111).

1185 1186 Personnel familiar with the equipment and typical site meteorological conditions should review 1187 the meteorological data. Data losses can be minimized by incorporating redundant sensors and 1188 equipment, and by maintaining an adequate inventory of spares, as part of the monitoring program design.

1189 Periodic data evaluation may include, but is not be limited to, promptly identifying and inspecting 1190 equipment failures and time to resolution, reviewing results of performance checks and calibrations, and 1191 confirming that measurements are within appropriate ranges (e.g., occurrence of excessive calm wind 1192 speeds, reasonable diurnal and seasonal variation of wind speed, wind direction, and temperature at each 1193 level and with height).

1194 1195 A change in /Q (and/or D/Q) may not be the only indicator that should be reviewed. A change 1196 in impact location should also be addressed (if not already the case). Such a change could be caused by 1197 (1) an actual change in the meteorological conditions, (2) a physical change in meteorological 1198 instrumentation (i.e., mechanical versus sonic anemometry), (3) a change in data averaging approach 1199 (e.g., scalar versus vector), or (4) any combination of the above.

1200 1201 Invalid data should be removed from the meteorological data file prior to calculating long-term, 1202 annual-average /Q and D/Q values. Records of data invalidation (and if applicable, data substitution) 1203 should also be documented and retained.

1204 1205 The long-term, annual-average /Q and D/Q values should be reevaluated periodically (e.g., every 1206 3-5 years). If the periodic reevaluation indicates the controlling/limiting long-term, annual-average /Q 1207 and D/Q values are substantially nonconservative (e.g., higher by 20-30 percent or more with respect to 1208 historical data), the licensee should ensure that the /Q and D/Q values used in the dose assessment are 1209 revised or that the ARERR addresses why such changes are not deemed necessary. Acceptable reasoning 1210 includes evaluating data anomalies, identification of failures in meteorological sensors, and 1211 documentation that the locality experienced abnormal weather patterns.

1212 1213 3.3 Release Height 1214 1215 The release height affects the dispersion (transport and diffusion) of radioactive materials, 1216 especially for downwash and building wake effects. For facilities with ground-level, mixed-mode, and 1217 elevated releases, an evaluation should be made to determine the proper location of the maximum 1218 exposed individual member of the public. From a dispersion perspective, when determining the 1219 maximum exposure location (submersion and/or deposition), the evaluation should consider the RG 1.21, Rev. 3, Page 27

1220 magnitude of the release(s) originating as an elevated release and as a ground-level release. For example, 1221 a close-in, downwind location in one sector may have a higher /Q (i.e., less dispersion) for a 1222 ground-level release, whereas the majority of the source term may be originating as an elevated release, 1223 causing a higher concentration () at a more distant location, possibly in a different sector. RG 1.111 1224 contains a more complete discussion of release height.

1225 1226 3.4 Aquatic Dispersion (Surface Waters) 1227 1228 Liquid radioactive effluents may be disposed in accordance with 10 CFR 20.2001 into a variety 1229 of receiving surface water bodies, including nontidal rivers, lakes, reservoirs, settling ponds, cooling 1230 ponds, estuaries, and open coastal waters. This effluent is dispersed by various mechanisms 1231 (i.e., turbulent mixing; stream flow in the water bodies; and internal circulation or flow-through in lakes, 1232 reservoirs, and cooling ponds). Parameters influencing the dispersion patterns and concentrations near a 1233 site include the direction and speed of flow of currents, both natural and plant induced, in the receiving 1234 water; the intensity of turbulent mixing; the size, geometry, and bottom topography of the receiving 1235 water; the location of effluent discharge in relation to the receiving water surface and shoreline; the 1236 amount of recirculation of previously discharged effluent; the characteristics of suspended and bottom 1237 sediments; and sediment sorption properties. RG 1.113 describes calculational models for estimating 1238 aquatic dispersion to surface water bodies. However, the dispersion characteristics may be highly site 1239 dependent, and local characteristics should be considered when performing dispersion modeling and dose 1240 assessments.

1241 1242 3.5 Spills and Leaks to the Ground Surface 1243 1244 Liquid releases onto the land surface are transported and diluted as a function of site-specific 1245 hydrologic features, events, and processes and properties of the effluent. The releases may temporarily 1246 accumulate, pool, or run off to natural or engineered drainage systems. During this process, water may 1247 also be absorbed into the soil (see Section 3.6). RG 1.113 discusses the use of simple models to estimate 1248 transport through surface water bodies and considers water usage effects. Spills or leaks of radioactive 1249 material to the ground surface should initiate characterization of the runoff. At a minimum, the 1250 characterization activities should satisfy (1) the requirements of 10 CFR 50.75(g) and (2) the effluent 1251 reporting requirements of 10 CFR 50.36a, and the guidance described in NUREG-1301 and 1252 NUREG-1302 for planned effluents (e.g., sampling before discharge to unrestricted areas).

1253 Sections 8.5.1, 8.5.2, and 8.5.9 of this RG contain recommendations on the general format for reporting 1254 abnormal releases to onsite areas and abnormal discharges to unrestricted areas.

1255 1256 3.6 Spills and Leaks to Groundwater 1257 1258 Liquid radioactive leaks and spills are sometimes released to onsite groundwater or discharged to 1259 offsite groundwater. Leaks and spills onto the ground surface can be absorbed into the soil. Depending 1260 on the local soil properties and associated liquid flux of the release, some of the material in the leak or 1261 spill may eventually reach the local water table. The dispersion of this material depends on the local 1262 subsurface geology and hydrogeologic characteristics. Liquid releases into the subsurface will be 1263 transported as a function of groundwater flow processes and conditions (e.g., hydraulic gradients, 1264 permeability, porosity, and geochemical processes) and will eventually be released to the unrestricted 1265 area.

1266 1267 A groundwater conceptual site model should be developed to predict the subsurface water flow 1268 parameters to include direction and rate and to be used as the basis for estimating the dispersion of 1269 abnormal releases of liquid effluents into groundwater (see RG 4.1 and RG 4.25). Section 1 of this RG 1270 lists references for use in developing an adequate groundwater conceptual site model.

RG 1.21, Rev. 3, Page 28

1271 1272 Simple analytical models or more rigorous numerical codes (i.e., simulations) may be used to 1273 evaluate subsurface transport following a release. Appropriate use of these models and codes will depend 1274 on the release rate, depth of the release, depth to the local water table, groundwater flow directions, 1275 groundwater flow rates, geochemical conditions, and other geochemical processes (e.g., geochemical 1276 retardation). Additionally, water usage, such as groundwater pumping from wells, may create local 1277 groundwater depression(s) that can alter the natural groundwater flow.

1278 1279 Consistent with 10 CFR 20.1501, a basic site hydrogeological characterization, in advance of 1280 leaks or spills, is helpful for evaluating potential leaks and spills. Sites with significant residual 1281 radioactivity that are likely to exceed the radiological criteria for unrestricted use at the time of 1282 decommissioning (e.g., as described in 10 CFR 20.1402) should perform more extensive evaluation.

1283 Initial assessments should be conducted with relatively simple conceptual site models using scoping 1284 surveys, bounding assumptions, or a combination of both (see RG 4.25 and American National Standards 1285 Institute/American Nuclear Society (ANSI/ANS) 2.17-2009, Evaluation of Subsurface Radionuclide 1286 Transport at Commercial Nuclear Power Production Facilities (Ref. 72). The complexity of the models 1287 should increase as (1) more knowledge is obtained about the system under evaluation (e.g., source of leak, 1288 plume size, concentrations, radionuclides, site characteristics, presence of preferential flow pathways) and 1289 (2) the dose estimates rise above significant residual radioactivity levels (see definition in the glossary).

1290 Industry documents (Refs. 38 and 72) contain details of various industry practices that may be used as 1291 part of a groundwater monitoring program. Sites with low-level spills or leaks generally do not require 1292 extensive site characterization and monitoring.

1293 1294 The following are basic steps in monitoring groundwater contamination:

1295 1296 1. Use the conceptual site model (as necessary) to assist in monitoring, evaluating, and reporting 1297 radioactive releases and radioactive discharges.

1298 1299 2. Collect empirical data by one or more of the following (as necessary):

1300 1301 a. Sample and analyze groundwater from existing monitoring wells.

1302 1303 b. Conduct additional hydrogeologic testing using existing wells (or new wells) if required.

1304 1305 3. Test the conceptual site model and radionuclide transport predictions using groundwater sample 1306 results and data collected during hydrogeologic testing.

1307 1308 4. Modify conceptual site model and radionuclide transport parameters as necessary to predict 1309 discharges and assess doses to members of the public.

1310 1311 5. Use an iterative process and revaluate as needed.

1312 1313 The groundwater monitoring results should be used in the development and testing of a 1314 conceptual site model to predict radionuclide transport in groundwater. The conceptual site model is 1315 generally considered adequate when it predicts the results of monitoring (sometimes called a calibrated 1316 model). Groundwater monitoring results evaluate the validity of the conceptual site model. Following a 1317 leak or spill of licensed (radioactive) material, the conceptual site model may be used in conjunction with 1318 radionuclide transport modeling and groundwater monitoring to comprise a basis for predicting future 1319 effluents from the site. Dispersion and dilution occur over time and in three dimensions.

1320 RG 1.21, Rev. 3, Page 29

1321 When used with a strategic and carefully planned monitoring program, the conceptual site model 1322 can ensure that necessary and reasonable surveys are performed (i.e., limited scoping surveys or more 1323 extensive surveys). Limited scoping surveys can determine if significant residual radioactivity exists and 1324 if there is adequate protection of public health and safety. If the limited scoping surveys identify 1325 significant residual radioactivity, then the extent of the contamination should be further evaluated by 1326 more extensive surveys (e.g., monitoring wells or other evaluations as appropriate). These survey 1327 activities may be direct (i.e., occurring at, or very near, the source of the leak) or indirect (i.e., occurring 1328 at some distance from the source of the leak) depending on the accessibility of the source of the spill or 1329 leak and the mobility of the radionuclides.

1330 1331 For spills or leaks occurring below the soil surface in inaccessible locations, direct scoping and 1332 characterization may not be feasible. In these cases, indirect monitoring techniques (e.g., groundwater 1333 monitoring wells in a down-gradient direction) will satisfy existing regulatory requirements. These 1334 survey activities should, at a minimum, satisfy (1) the requirements of 10 CFR 50.75(g) and (2) the 1335 effluent reporting requirements of 10 CFR 50.36a for groundwater discharges to the unrestricted area. In 1336 general, licensees should describe (report) leaks and spills of radioactive material in the ARERR for the 1337 calendar year the spill or leak occurred. Additionally, licensees should report groundwater monitoring 1338 data in the ARERR for the calendar year in which the data were collected. Sections 8.5.1, 8.5.2, and 8.5.9 1339 of this RG contain guidance on the general format for reporting abnormal releases to onsite areas and 1340 abnormal discharges to unrestricted areas.

1341 1342 Although licensees may conduct a groundwater monitoring effort for different reasons, for 1343 purposes of this RG, the surveys, characterization activities, conceptual site models, and other 1344 components of any groundwater monitoring effort should be sufficient to do the following:

1345 1346 1. Appropriately report, for purposes of accountability, effluents discharged to unrestricted areas.

1347 1348 2. Document information in a format consistent with Table A-6 and Section 8.5 of this RG.

1349 1350 3. Provide advance indication of potential future discharges to unrestricted areas (to ensure releases 1351 are planned and monitored before discharge).

1352 1353 4. Demonstrate that significant residual radioactivity has not migrated off site to an unrestricted area 1354 in the annual reporting interval.

1355 1356 5. Communicate relevant information as described in Section 9.5 of this guide.

1357 1358 4. Quality Assurance 1359 1360 4.1 Quality Assurance Programs 1361 1362 The analytical process should use a range of QA checks and tests. RG 4.15 describes the QA 1363 program activities for ensuring that radioactive effluent monitoring systems and operational programs 1364 meet their intended purpose. Each licensees licensing basis determines the applicability of Revision 1 or 1365 Revision 2. However, RG 4.15, Revision 2 contains guidance on determining appropriate sensitivity 1366 levels for analytical instrumentation based on DQOs. The use of DQOs may provide a better technical 1367 basis for determining sensitivity levels (e.g., LLDs) than the use of the default values in NUREG-1301 1368 and NUREG-1302. A combination approach using both Revision 1 and Revision 2 of RG 4.15 may be 1369 used to determine appropriate sensitivity levels (e.g., LLDs) different (i.e., higher or numerically larger) 1370 than those listed in NUREG-1301 and NUREG-1302.

1371 RG 1.21, Rev. 3, Page 30

1372 4.2 Quality Control Checks 1373 1374 QC checks of laboratory instrumentation should be conducted daily or before use, and 1375 background variations should be monitored at regular intervals to demonstrate that a given instrument is 1376 in working condition and functioning properly. QC records should include results of routine tests and 1377 checks, background data, calibrations, and all routine maintenance and service.

1378 1379 4.3 Surveillance Frequencies 1380 1381 Routine qualitative tests and checks (e.g., channel operational tests, channel checks, or source 1382 checks to demonstrate that a given instrument is in working condition and functioning properly) may be 1383 performed using radioactive sources that are not traceable by the National Institute of Standards and 1384 Technology (NIST). The schedule for source checks, channel checks, channel calibrations, and channel 1385 operational tests should be in accordance with NUREG-1301 and NUREG-1302, unless otherwise 1386 modified after a technical evaluation demonstrates a justifiable change in frequency. A technical 1387 evaluation that revises a surveillance frequency should include consideration of the instruments function 1388 and the consequences of failure and not simply rely on the history of successful surveillances.

1389 1390 4.4 Procedures 1391 1392 Individual written procedures should be used to establish specific methods of calibrating installed 1393 radiological monitoring systems and grab sampling equipment. Written procedures should document 1394 calibration practices used for ancillary equipment and systems (e.g., meteorological equipment, airflow 1395 measuring equipment, in-stack monitoring pitot tubes). Calibration procedures may be compilations of 1396 published standard practices or manufacturers instructions that accompany purchased equipment, or they 1397 may be written in house to include special methods or items of equipment not covered elsewhere.

1398 Calibration procedures should identify the specific equipment or group of instruments to which the 1399 procedures apply.

1400 1401 Written procedures should be used for maintaining counting room instrument accuracy, including 1402 maintenance, storage, and use of radioactive reference standards; instrumentation calibration methods; 1403 and QC activities such as collection, reduction, evaluation, and reporting of QC data as required by the 1404 technical specifications.

1405 1406 4.5 Calibration of Laboratory Equipment and Routine Effluent Radiation Monitors 1407 1408 Calibrations (e.g., of laboratory equipment and continuous radiation monitoring systems used to 1409 quantify radioactive effluents) should be performed using the general principles for calibration of effluent 1410 monitoring instrumentation provided in ANSI N42.18-2004 and ANSI N323C-2009, American National 1411 Standard for Radiation Protection Instrumentation Test and CalibrationAir Monitoring Instruments, 1412 American National Standards Institute (Ref. 74), using radioactive calibration sources traceable to the 1413 NIST. Calibration sources should have the necessary accuracy, stability, and radioactivity levels required 1414 for their intended use. The relationship between concentrations and monitor readings should be 1415 determined. Performance of the monitoring system should be judged on the basis of reproducibility, time 1416 stability, and sensitivity.

1417 1418 Periodic inservice correlations that relate monitor readings to the concentrations, release rates of 1419 radioactive material in the monitored release path, or a combination of both, should be performed when 1420 possible to validate the adequacy of the system. These correlations should be based on the results of 1421 analyses for specific radionuclides in grab samples from the release path.

1422 RG 1.21, Rev. 3, Page 31

1423 The use of NIST-traceable sources combined with mathematical efficiency calibrations may be 1424 applied to instrumentation used for radiochemical analysis (e.g., gamma spectroscopy systems) if 1425 employing a method provided by the instrument manufacturer.

1426 1427 4.6 Calibration of Measuring and Test Equipment 1428 1429 Measuring and test equipment should be calibrated using NIST-traceable radioactive sources.

1430 The source geometries should be representative of the sample types analyzed and have the necessary 1431 accuracy, stability, and activity concentrations for their intended use.

1432 1433 4.7 Calibration Frequency 1434 1435 Calibrations should generally be performed at regular intervals in accordance with the frequencies 1436 established in NUREG-1301 and NUREG-1302. A change in calibration frequency (an increase or 1437 decrease) should be based on the reproducibility and time stability characteristics of the system. For 1438 example, an instrument system that gives a relatively wide range of readings when calibrated against a 1439 given standard should be recalibrated at more frequent intervals than one that gives measurements within 1440 a more-narrow range. Any monitoring system or individual measuring equipment should be recalibrated 1441 or replaced whenever it is suspected of being out of adjustment, excessively worn, or otherwise damaged 1442 and not operating properly.

1443 1444 4.8 Measurement Uncertainty 1445 1446 The measurement uncertainty (formerly called measurement error) associated with the 1447 measurement of radioactive materials in effluents should be estimated. Counting statistics can provide an 1448 estimate of the statistical counting uncertainty involved in radioactivity analyses. Because it may be 1449 difficult to assign error terms for each parameter affecting the final measurement, detailed statistical 1450 evaluations of error are not required. Normally, the statistical counting uncertainty decreases as the 1451 amount (concentration) of radioactivity increases. Thus, for the radioactive effluent release report, the 1452 statistical counting uncertainty is typically a small component of the total uncertainty. The sampling 1453 uncertainty is likely the largest component and includes uncertainties such as the uncertainty in 1454 volumetric and flow-rate measurements and laboratory processing uncertainties.

1455 1456 The total or expanded measurement uncertainty associated with the effluent measurement should 1457 ideally include the cumulative uncertainties resulting from the total operation of sampling and 1458 measurement. Expanded uncertainty should be reported with measurement results. The objective should 1459 be to evaluate only the important contributors and obtain a reasonable measure of the uncertainty 1460 associated with reported results. Detailed statistical and experimental evaluations are not required. The 1461 overall objective should be to obtain an overall estimate of measurement uncertainty. The formula for 1462 calculating the total or expanded uncertainty classically includes the square root of the sum of squares of 1463 each important contributor to the measurement uncertainty. Licensees may obtain additional information 1464 from NUREG-1576 and ANSI/HPS N13.1-2011.

1465 1466 4.9 Calibration of Accident-Range Radiation Monitors and Accident-Range Effluent Monitors 1467 1468 GDC 64 requires means for monitoring radioactivity in the reactor containment atmosphere; 1469 spaces containing components for recirculation of loss-of-coolant accident (LOCA) fluids; effluent 1470 discharge paths; and the plant environs for radioactivity that may be released from normal operations, 1471 including anticipated operational occurrences, and from postulated accidents. The regulation at 1472 10 CFR 20.1501(c) requires periodic calibration of instruments and equipment used to perform 1473 quantitative radiation measurements (e.g., dose rate and effluent monitoring).

RG 1.21, Rev. 3, Page 32

1474 1475 NUREG-0737, Item II.F.1, provides guidance for monitoring radiation levels and gaseous 1476 effluent during postulated radiological emergencies. RG 1.97, Revisions 2 and 3, provide guidance on the 1477 design and performance criteria of instrumentation used to assess plant and environ conditions during and 1478 following an accident. This RG 1.21 provides further guidance on the calibration of such instrumentation 1479 based on the NRCs Proposed Guidance for Calibration and Surveillance Requirements to Meet Item 1480 II.F.1 of NUREG-0737, issued August 1982 (Ref. 75). NUREG/CR-5569, Health Physics Positions 1481 Data Base, Health Physics Position (HPPOS)-001, Proposed Guidance for Calibration and Surveillance 1482 Requirements to Meet Item II.F.1 of NUREG-0737, issued February 1994 (Ref. 76), summarizes this 1483 additional guidance.

1484 1485 Noble Gas Monitoring - NUREG-0737, Item II.F.1-1, describes accident-range noble gas effluent 1486 monitors as monitors that are normally noble gas gross activity monitors sensitive to gamma emissions, 1487 beta emissions, or a mix of gamma and beta emissions. These monitors normally indicate (read out) in 1488 units of activity concentration, a count rate, or a dose rate (i.e., an indirect measurement of the noble gas 1489 gross activity concentration). Therefore, in order to determine the release rate of noble gas gross activity, 1490 a conversion factor (i.e., hereafter referred to as an instrument response factor) should be developed to 1491 convert the instrument output into an activity concentration for use in determining a release rate 1492 (e.g., curies per second of a mix of noble gases).

1493 1494 The initial vendor calibration of emergency effluent monitoring instruments may be a one-time 1495 prototype calibration based on the initial calibration of a single instrument of a certain model using 1496 NIST-traceable radiation sources. This initial prototype calibration of a single instrument model may 1497 include a determination of its fundamental characteristics, such as the following:

1498 1499 1. a dose-rate linearity check using a radioactive gas or solid source (e.g., cesium (Cs)-137) to 1500 obtain three on-scale values separated by two decades of scale; 1501 1502 2. a measurement of the instruments response factor to a calibration gas (e.g., xenon (Xe)-133 or 1503 krypton (Kr)-85);

1504 1505 3. a characterization of the instruments energy -dependency characteristics, using solid sources 1506 ranging in gamma energy from low energy (e.g., 81 kiloelectron volts) to high energy 1507 (e.g., 2 megaelectron volts); and 1508 1509 4. a determination, using a solid source, of a transfer factor that provides a dual purpose:

1510 1511 a. for use by vendors to validate that subsequent instruments produced for sale of the same 1512 model have similar performance characteristics to the initial type instrument models 1513 characteristics; and 1514 1515 b. for use by end users (e.g., nuclear power plants) in performing post installation and 1516 subsequent periodic calibration to verify that the instruments installed in the facility are 1517 functioning consistently with respect to initial vendor calibration of that instrument 1518 model.

1519 1520 Time-dependent (i.e., time since reactor shutdown) instrument response factors may be developed 1521 for each major accident type (i.e., a small-break LOCA with normal reactor coolant system activity levels, 1522 a large-break LOCA with gas gap activity levels, or a core-melt accident with noble gas activity levels 1523 arising from the fuel pellets release of noble gas). Each accident type has a characteristic, time-dependent RG 1.21, Rev. 3, Page 33

1524 noble gas isotopic mix. In general terms, a small-break LOCA has a substantially decayed noble gas mix 1525 from the reactor coolant system with predominantly low-energy gamma photons; a large-break LOCA has 1526 a somewhat decayed noble gas mix from the gas gap of the fuel assemblies with predominately medium-1527 energy gamma photons; and a core -melt accident has a substantially undecayed mix of noble gas isotopes 1528 in the fuel pellets with predominately high-energy gamma photons.

1529 1530 The time-dependent instrument response factor accounts for the detectors energy efficiency at 1531 various gamma energies of the noble gas isotopic mix for that accident type. The instrument response 1532 factor normally has units of microcuries per cubic centimeter (µCi/cc) per count per minute or µCi/cc per 1533 milliroentgen per hour where the µCi/cc is the gross (total) summation of all the noble gas activities in the 1534 isotopic mix for each major type of accident listed above. It is also acceptable to use instrument response 1535 factors based on a single calibration gas with a low -energy gamma source (e.g., Xe-133) or beta 1536 emissions (e.g., Kr-85) for beta sensitive monitors.

1537 1538 The initial calibration process performed by the vendor does not need to be repeated at a nuclear 1539 power plant. Instead, a periodic single point source response check of the instruments performance as 1540 compared to a transfer factor provided by the vendor using a solid source - see ANSI N320-1978, 1541 Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation (Ref. 77).

1542 1543 Iodine and Particulate Monitoring - NUREG-0737, Item II.F.1-2, provides guidance on iodine 1544 and particulate effluent monitoring by sampling and analysis. Real-time monitoring is not required or 1545 considered practical; however, the licensees should have established procedures for collection of iodine 1546 and particulate samples and subsequent analysis to determine the release rate. For emergency dose 1547 assessment purposes, RASCAL (NUREG-1940 Section 1.2.8) can also be used to assess a real-time 1548 iodine and particulate release rate based on partitioning (scaling) factors to noble gases.

1549 1550 Containment High Range Monitoring - NUREG-0737, Item II.F.1-3, provides guidance on 1551 calibration of containment high-range monitors. An in-place calibration should be performed using a 1552 radioactive source at one point on the decade below 10 roentgens per hour (R/hr). Instrument scales in 1553 the range of 10 R/hr to 1E7 R/hr should be checked using electronic signal substitution with a calibrated 1554 current source to demonstrate that the system is functioning to higher radiation fields.

1555 1556 Containment high-range monitors should be used to assess the amount of core damage and to 1557 assess the source terms for the containment leakage release pathway. NUREG-1940, Section 1.2.4, 1558 Figures 1-1 through 1-5, provide information for PWRs and BWRs at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor 1559 shutdown that correlates the containment radiation monitor readings to the amount of reactor damage for 1560 normal coolant, spiked coolant, cladding failure, and core melt accident scenarios.

1561 1562 5. Dose Assessments for Individual Members of the Public 1563 1564 The regulation in 10 CFR 20.1301 establishes dose limits for individual members of the public8.

1565 The regulations referenced in Sections 5.4-5.6 of this RG contain both dose limits and design objectives 1566 that the licensee demonstrates compliance with through calculations. Table 1 summarizes the 1567 fundamental parameters associated with the dose calculations. RG Sections 5.7 and 5.8 present important 1568 concepts for these calculations. Because of differences between NRC and EPA regulations, 8

For ISFSIs, 10 CFR Part 72 specifies dose limits for any real individual beyond the Part 72 controlled area boundary (excluding occupational exposures). Thus, dose assessments performed to demonstrate compliance with the 10 CFR 72.104 must include the necessary components described in10 CFR 72.104.

RG 1.21, Rev. 3, Page 34

1569 demonstrating compliance only with radiological effluent technical specifications (based on 1570 10 CFR Part 50, Appendix I) does not necessarily ensure compliance with the EPAs 40 CFR Part 190, 1571 particularly if there is a direct radiation component (e.g., from BWR shine, ISFSI, or radioactive materials 1572 storage).

1573 1574 Table 1 - Parameters Associated with Dose Calculations 1575 10 CFR 20.1301(e) 10 CFR PART 50, APPENDIX I (EPA 40 CFR PART 190)

Dose whole body, max of any organ, whole body, thyroid, and max of any organ gamma air, and beta air Basis International Commission on EPA 40 CFR Part 190 Radiation Protection (ICRP)-2, Report of Committee II on Permissible Dose for Internal Radiation, issued 1959 (Ref. 78)

Where unrestricted area unrestricted area Individual real person/exposure pathway real person/exposure pathway (nearest real Receptor (nearest real residence, real garden, (residence, real garden, real dairy/meat real dairy/meat animal) animal)

Origin liquid and gas radioactive waste liquid and gas radioactive waste direct radiation (e.g., nitrogen-16 shine, ISFSI, radioactive materials storage, outside tanks) accumulated radioactive material (e.g., tritium in lake water) not already included in dose estimates Radioactive licensed only (per Appendix I, licensed and unlicensed Material Section II radioactive materials - see (see Section 5.6 below)

Section 5.4 below)

When current year current and prior years operation 1576 1577 5.1 Bounding Assessments 1578 1579 Bounding assessments may be useful if compliance can be readily demonstrated using 1580 conservative assumptions. In this RG, the term bounding assessment means that the reported value is 1581 unlikely to be substantially underestimated (see 10 CFR Part 50, Appendix I, Section III). Bounding 1582 assessments for the current year do not imply the absolute bounds for future conditions.

1583 1584 For example, licensees may use conservative bounding dose assessments in lieu of site-specific 1585 dose assessments of the maximum dose to individual members of the public. Instead of assessing dose 1586 from ground-level effluent releases to a real individual member of the public located 3.2 km (2 miles) 1587 from the site boundary, a conservative bounding dose assessment can be performed for a hypothetical 1588 individual member of the public located at the site boundary.

1589 1590 If bounding assumptions are made, the radioactive effluent release report should state such and 1591 should annotate the assumptions. Hypothetical exposure pathways (see definition in the glossary) and 1592 locations are sometimes used for bounding dose assessments (or hazard evaluations done in accordance 1593 with 10 CFR 20.1501).

1594 RG 1.21, Rev. 3, Page 35

1595 5.2 Individual Members of the Public 1596 1597 Individual members of the public reside in the unrestricted area but at times may enter the 1598 controlled area of a commercial nuclear power plant. Each licensee is responsible for classifying 1599 individuals (by location) as either members of the public or as occupational workers (see the definition of 1600 members of the public in 10 CFR Part 20.) The NRC annual dose limits for members of the public are 1601 100 mrem total effective dose equivalent in accordance with 10 CFR 20.1301(a) and (b). In addition, in 1602 accordance with 10 CFR 20.1301(e) for uranium fuel cycle licensees (including nuclear power plants), the 1603 annual dose limits to members of the public are the EPA 40 CFR Part 190 limits of 25 mrem whole body, 1604 75 mrem to the thyroid, and 25 mrem to any other organ while in the unrestricted area.

1605 1606 If bounding assessments are not used, licensees should perform evaluations to determine the dose 1607 to a real, maximum exposed member of the public, regardless of whether the individual is in an 1608 unrestricted area or a controlled area. A member of the public is typically a real individual in a 1609 designated location where there is a real exposure pathway (e.g., a real garden, real cow, real goat, or 1610 actual drinking water supply) and not a fictitious fencepost resident or an exposure pathway that includes 1611 a virtual goat or cow. Licensees are encouraged (but not required) to use real individual members of the 1612 public when performing dose assessments for radioactive discharges. Table 1 in RG 1.109 allows a dose 1613 evaluation to be performed at a location where an exposure pathway and dose receptor actually existed at 1614 the time of licensing.

1615 1616 5.3 Occupancy Factors 1617 1618 For members of the public in the unrestricted area, occupancy factors should be assumed to be 1619 100 percent at locations identified in the land use census, unless site -specific information indicates 1620 otherwise. Occupancy factors may be applied inside the controlled area based on estimated hours spent in 1621 the controlled area.

1622 1623 5.4 10 CFR Part 50, Appendix I, Design Objectives and Limiting Conditions for Operation 1624 1625 Appendix I to 10 CFR Part 50 contains numerical guidance for design objectives and limiting 1626 conditions of operation for radioactive waste systems to ensure discharges of radioactive liquid and 1627 gaseous effluents to unrestricted areas are ALARA. This numerical guidance is listed in terms of annual 1628 air doses (gamma and beta), annual total body doses, and annual organ doses (see below). Licensee 1629 technical specifications require that exposure to liquid and gaseous effluents conform to the numerical 1630 guidance in 10 CFR Part 50, Appendix I. In accordance with 10 CFR 50.34a, these numerical guides for 1631 design objectives and limiting conditions of operation are not to be construed as radiation protection 1632 standards. For these dose calculations, the following terms are generally used:

1633 1634 1. air doses (gamma and beta), total body doses, and organ doses (based on ICRP-2),

1635 2. effluent discharges only (excludes direct radiation from the facility and ISFSIs),

1636 3. current annual period (excludes accumulated radioactivity from prior-year effluents), and 1637 4. unrestricted area (excludes individuals in the restricted areas and controlled areas).

1638 1639 When calculating air doses, licensees should assure that, for any location outside the site 1640 boundary, doses do not exceed the dose objectives in 10 CFR Part 50, Appendix I. Calculation of air 1641 dose at the site boundary would assure the most conservative calculation of air doses for ground-level 1642 releases. This may not be true for elevated releases. Licensees should select a location that assures the 1643 most conservative calculation of air dose.

1644 RG 1.21, Rev. 3, Page 36

1645 5.5 10 CFR 20.1301(a) NRC dose limits for individual members of the public 1646 1647 This regulation specifies dose limits for members of the public from licensed operation of the 1648 facility. These limits apply to doses resulting from licensed and unlicensed radioactive material and from 1649 radiation sources other than background radiation (see 10 CFR 20.1001, Purpose). The dose limits 1650 include contributions to doses from (1) current-year effluents, (2) current-year direct radiation from the 1651 facility, and (3) accumulated radioactivity from prior-year effluents. The Technical Specifications 1652 establish the Radioactive Effluent Controls Program and the Environmental Monitoring Program, which 1653 establish effluent control methods sufficient to demonstrate of compliance with the NRC public dose 1654 limits in 10 CFR 20.1301(a).

1655 1656 5.6 10 CFR 20.1301(e) EPA Environmental Radiation Standards for the Uranium Fuel Cycle 1657 1658 For those facilities subject to the EPAs generally applicable environmental radiation standards in 1659 40 CFR Part 190, licensees must assess the highest cumulative (whole body and organ) doses from the 1660 uranium fuel cycle to a real individual in the general environment (i.e., outside the site boundary). The 1661 dose limits include contributions to doses from (1) current-year effluents, (2) current-year direct radiation 1662 from the facility, and (3) accumulated radioactivity from prior-year effluents. The Technical 1663 Specifications establish the Radioactive Effluent Controls Program and the Environmental Monitoring 1664 Program, which establish effluent control requirements sufficient to demonstrate compliance with the 1665 EPA public dose limits in 40 CFR Part 190 (see NUREG-0543).

1666 1667 These requirements include the following considerations:

1668 1669 1. Whole body and organ doses come from ICRP-2 concepts.

1670 1671 2. Any member of the public means any individual except when that individual is receiving an 1672 occupational dose.

1673 1674 3. The unrestricted area means an area, access to which is neither limited nor controlled by the 1675 licensee. The boundaries of the unrestricted area are defined by the licensee. (See also the 1676 definition of generally applicable environmental radiation standards in 10 CFR 20.1003.)

1677 1678 4. Current-year effluents includes both normal and abnormal discharges to the unrestricted area.

1679 1680 5. Current-year direct radiation includes all direct radiation from the facility (e.g., radioactive 1681 waste storage and ISFSIs) but excludes doses from radioactive waste shipments.

1682 1683 6. Cumulative dose means the sum of (1) current-year effluent dose, (2) current-year direct 1684 radiation dose, and (3) dose from accumulated radioactivity if not already included in the first two 1685 items.

1686 1687 7. Accumulated radioactivity includes radioactive material in the unrestricted area from prior-year 1688 discharges that remains in the environment (e.g., tritium in lake water or radionuclides).

1689 1690 8. The uranium fuel cycle excludes uranium mining, radioactive waste shipping (in the 1691 unrestricted area), operations at waste disposal sites, and reuse of nonuranium special nuclear 1692 materials. (See the definition of uranium fuel cycle in 40 CFR Part 190 and in the glossary of 1693 this document.)

1694 RG 1.21, Rev. 3, Page 37

1695 5.7 Dose Assessments for 10 CFR Part 50, Appendix I 1696 1697 Dose assessments to show compliance with technical specification requirements for meeting the 1698 numerical values of 10 CFR Part 50, Appendix I, design objectives should include quarterly and annual 1699 doses using the considerations in Section 5.4 of this RG. The dose assessments should be reported in a 1700 format similar to that shown in Table A-4 in Appendix A to this RG and include the items listed below:

1701 1702 1. doses from liquid effluents 1703 a. total body dose, quarterly and annually; 1704 b. organ dose, quarterly and annually (maximum, any organ); and 1705 c. percent of limits for each of the above.

1706 1707 2. doses from gaseous effluents 1708 a. beta and gamma air doses, quarterly and annually; 1709 b. organ dose commitment from iodine, tritium, and particulate releases with half-lives 1710 greater than 8 days, quarterly and annually; and 1711 c. percent of limit for each of the above.

1712 1713 An evaluation of the local exposure pathways to determine the maximum exposed member of the 1714 public should be performed. However, maximum doses from various exposure pathways are not additive 1715 from different locations. For example, dose from a downstream drinking water exposure pathway should 1716 not be added to the dose to an upstream resident whose exposure is from gaseous effluents and direct 1717 radiation unless that individuals drinking water is obtained from the downstream location.

1718 1719 Maximum doses to real individuals should be assessed as described in RG 1.109. The locations 1720 and exposure pathways are those where real individuals are present and exposed. Maximum exposed 1721 individuals are characterized as maximum with regard to food consumption, occupancy, and other 1722 usage in the vicinity of the plant site. For example, licensees should make maximum assumptions for 1723 food consumption and occupancy factors at actual locations when assessing dose to the maximum 1724 exposed individual, unless they have determined and applied site -specific (actual) data. In lieu of 1725 assessing dose to real individuals, licensee may also use bounding dose assessments for compliance with 1726 10 CFR Part 50, Appendix I (see the section titled Bounding Assessments).

1727 1728 The objective of 10 CFR Part 50, Appendix I, is to provide numerical guides for design objectives 1729 and limiting conditions for operation to ensure that radioactive effluent control equipment is effective in 1730 reducing emissions to ALARA levels. The numerical guidance pertains to quarterly and annual dose 1731 criteria at or beyond the unrestricted area from current-year effluent discharges. The calculations related 1732 to Appendix I do not include dose from radioactivity in prior-year, accumulated, effluent discharges 1733 (e.g., last years radioactivity remaining in lake water is excluded). However, the dose calculations for 1734 demonstrating compliance with the EPA limits do include accumulated radioactivity (see Section 5.8 of 1735 this RG).

1736 1737 For purposes of demonstrating compliance with dose criteria for limiting dose to a member of the 1738 public in unrestricted areas in accordance with Technical Specifications conforming to 10 CFR 50, 1739 Appendix I, the exposure pathways and routes of exposure identified in RG 1.109 should be considered.

1740 An evaluation of other exposure pathways (not included in dose assessments) should be performed and 1741 maintained for purposes of demonstrating compliance with the staff position on significant exposure 1742 pathways. Calculational procedures should be based on models and data such that the actual exposure of 1743 an individual through appropriate pathways is unlikely to be substantially underestimated. A new 1744 exposure pathway should be included in the demonstration of compliance if the calculated dose from that 1745 new exposure pathway exceeds 10 percent of the 10 CFR 50 Appendix I, Section II numerical guides on RG 1.21, Rev. 3, Page 38

1746 design objectives. Bounding dose assessments as described in Section 5.1 of this RG may be used in 1747 evaluating the dose from any new significant exposure pathways.

1748 1749 Real exposure pathways are identified for routine discharges and direct radiation based on the 1750 results of the land use census. Dose calculations should typically be performed based on real exposure 1751 pathways. Conversely, dose assessments (i.e., surveillances and dose calculations) are not needed for 1752 exposure pathways that do not exist at a site. For example, if the land use census does not identify the 1753 existence of an ingestion exposure pathway involving a milk animal, the licensee is not required to assess 1754 that route of exposure for the ingestion exposure pathway. Similarly, if a licensee discharges liquid 1755 radioactive waste to a body of water (either surface water or groundwater) and that body of water is not 1756 used as a source of drinking water (either private or public), a drinking water assessment is not required.

1757 For purposes of reporting information in the ARERR, there is a distinction between dose assessments for 1758 10 CFR Part 50, Appendix I, and hazard assessments that may be conducted for onsite spills and leaks, as 1759 outlined in 10 CFR 20.1501 (where bounding estimates may be necessary). (See the discussion of 1760 bounding dose estimates in Section 5.1 of this RG.)

1761 1762 5.8 Dose Assessments for 10 CFR 20.1301(e) 1763 1764 To show compliance with 10 CFR 20.1301(e), dose assessments should be reported according to 1765 the generally applicable environmental radiation standards in 40 CFR Part 190, with consideration of 1766 Section 5.6 of this RG, and in a format similar to Table A-5 of Appendix A to this RG.

1767 1768 5.8.1 The following should be reported:

1769 1770

• whole body dose to the maximum individual member of the public, 1771

• thyroid dose to the maximum individual member of the public, 1772

• dose to any other organ to the maximum individual member of the public, and 1773

• percent of the applicable limit.

1774 1775 5.8.2 One means of demonstrating compliance with 40 CFR Part 190 is listed in Volume 42 of the 1776 Federal Register, page 2859, which states the following:

1777 1778 In the case of light water reactors, demonstrating conformance with 1779 Appendix I of 10 CFR 50 are generally adequate for demonstrating compliance 1780 with [EPA 40 CFR Part 190].

1781 1782 As a result, a licensee that (1) can demonstrate that external sources of direct radiation are 1783 indistinguishable from background and (2) demonstrates compliance with the numerical dose guidance of 1784 10 CFR Part 50, Appendix I, may cite the above reference as the basis for demonstrating compliance with 1785 40 CFR Part 190. The NRC provides additional guidance in NUREG-0543, Methods for Demonstrating 1786 Compliance with the EPA Uranium Fuel Cycle Standard (40 CFR Part 190).

1787 1788 However, licensees that (1) have external sources of direct radiation that are above background 1789 and (2) demonstrate compliance with the numerical dose guidance of 10 CFR Part 50, Appendix I, must 1790 also include sources of direct radiation from uranium fuel cycle operations (e.g., including direct radiation 1791 from the licensed facility and co-located or nearby nuclear power facilities, as appropriate).

1792 1793 5.8.3 The dose contributions from direct radiation may be estimated based on either (1) direct radiation 1794 measurements (e.g., thermoluminescent dosimeters, optically stimulated dosimeters, radiation 1795 detection instruments), (2) calculations, or (3) a combination of measurements and calculations.

RG 1.21, Rev. 3, Page 39

1796 When direct radiation dose is determined by measurement, RG 4.13 provides guidance on 1797 determining the dose to members of the public. Several sources contain additional information 1798 on background subtraction for environmental dosimeters (Refs. 29, 79, 80, and 81). Methods of 1799 determining dose from direct radiation to the maximum exposed individual member of the public 1800 may also include extrapolation methods.

1801 1802 Licensees must demonstrate compliance with 10 CFR 20.1301(e) for the generally applicable 1803 environmental radiation standards in 40 CFR Part 190. These include the concept of a total dose (to the 1804 whole body and to any organ) from all sources related to the uranium fuel cycle (such as adjacent or 1805 nearby nuclear power plants).

1806 1807 Contributions to the total dose from radioactive effluents (liquid and gaseous) and direct radiation 1808 should be included, if applicable. Other sources (e.g., accumulated radioactive materials in offsite ponds 1809 or lakes from previous years discharges) should also be included, if applicable, when estimating the total 1810 dose. However, if the contributions from direct radiation or accumulated radioactivity are generally 1811 minor (as evaluated and documented in a licensee technical evaluation as not contributing to the total 1812 dose), these contributions need not be included in the total dose evaluation, but the basis for exclusion 1813 should be documented.

1814 1815 5.9 Dose Calculations 1816 1817 Acceptable dose assessment models, such as those provided in RGs 1.109, 1.111, 1.112, and 1818 1.113, should be used to make dose calculations. When calculating organ doses from airborne effluents 1819 for purposes of demonstrating compliance with Technical Specifications conforming to 10 CFR 50, 1820 Appendix I, contributions from I-131, I-133, tritium, and radionuclides in particulate form with half-lives 1821 greater than 8 days should be included in the assessment. For demonstrating compliance with EPA 40 1822 CFR 190, in addition to the above radionuclides, doses from C-14 should be included in organ dose 1823 assessments.

1824 1825 6. Solid Radioactive Waste Released from the Unit 1826 1827 Section 5.6, Reporting Requirements, in the Standard Technical Specifications normally 1828 requires reporting of solid waste released from the unit (see NUREG-1430, 1431, 1432, 1433, and 1434 1829 (Refs. 82 - 86)). The data reported should be for the LLW volumes shipped from the unit (plant site).

1830 1831 Solid radioactive waste shipments should be reported in a format similar to that of Table A-3 in 1832 Appendix A to this RG. The total curie quantity and major radionuclides in the solid waste shipped off 1833 site should be determined and reported.

1834 1835 The data should be divided by the waste stream categories listed in Table A-3. The waste streams 1836 are:

1837 (1) wet radioactive waste (e.g., spent resin, filters, sludges, etc.),

1838 (2) dry radioactive waste (e.g., trash, paper, discarded protective clothing etc.),

1839 (3) activated or contaminated radioactive material (e.g., equipment or bulk radioactive material, 1840 etc.), and 1841 (4) other radioactive waste (waste not included in the above categories and not excepted from 1842 reporting as described below).

1843 1844 Shipments that do not need to be reported include shipments of metal melt, contaminated 1845 equipment for transfer between licensees or equipment for refurbishment, contaminated laundry (either 1846 launderable or dissolvable), or radioactive samples for analysis. Potentially contaminated dry active RG 1.21, Rev. 3, Page 40

1847 waste sent for resurvey and segregation (sometimes referred to as green is clean) does not need to be 1848 reported. Equipment shipped for decontamination and free release does not need to be reported.

1849 However, records of these types of shipments should be maintained on site.

1850 1851 Note 1: Data on LLW disposed in licensed LLW disposal facilities is available using the Manifest 1852 Information Management System operated by the U.S. Department of Energy.

1853 1854 Note 2: There are no requirements for reporting storage of LLW at nuclear power plants.

1855 However, LLW storage records should be established and maintained at nuclear plants and made 1856 available for NRC inspection during routine effluent inspections consistent with applicable NRC 1857 requirements.

1858 1859 1860 7. Reporting Errata in Effluent Release Reports 1861 1862 Errors in radioactive effluent release reports should be classified and reported as described below.

1863 1864 7.1 Examples of Small Errors 1865 1866 Small errors may be any of the following:

1867 1868 1. inaccurate reporting of dose that equates to 10 percent of the applicable 10 CFR Part 50, 1869 Appendix I, design objective or 10 percent of the EPA public dose criterion; 1870 1871 2. inaccurate reporting of curies (or release rates, volumes, etc.) that equate to 10 percent of the 1872 affected curie total (or release rate, volume, etc.) after correction; 1873 1874 3. omissions that do not impede the NRCs ability to adequately assess the information supplied by 1875 the licensee; or 1876 1877 4. typographical errors or other errors that do not alter the intent of the report.

1878 1879 7.2 Reporting Small Errors 1880 1881 Licensees should correct small errors within 1 year of discovery and may submit the correction 1882 with the next (normally scheduled) submittal of the ARERR, as follows. A brief narrative explanation of 1883 the errors should be included in Section 8, Errata/Corrections to Previous ARERRs, of Table A-6. The 1884 narrative should state that the affected pages, in their entirety, are included as attachments to the ARERR.

1885 Additionally, the corrected pages, in their entirety, should be submitted as an attachment (or addendum) 1886 to the ARERR. The corrected pages should reference the affected calendar year and should contain 1887 revision bars in the margins of the page to indicate the locations of the changes. If submitting corrections 1888 to multiple ARERRs, a separate attachment (or addendum) should be made for each of the affected years.

1889 Other methods of correcting previous ARERRs may be used, provided the corrections are clearly and 1890 completely described.

1891 1892 7.3 Examples of Large Errors 1893 1894 Large errors may be any of the following:

1895 1896 1. inaccurate reporting of dose that equates to >10 percent of the 10 CFR Part 50, Appendix I, or 1897 EPA public dose criterion, after correction; RG 1.21, Rev. 3, Page 41

1898 1899 2. inaccurate reporting of curies (or release rate, volume, etc.) that equates to >10 percent of the 1900 affected curie total (or release rate, volume, etc.) after correction; 1901 1902 3. omissions that may impede the NRCs ability to adequately assess the information supplied by 1903 the licensee; or 1904 1905 4. typographical errors or other errors that significantly alter the intent of the report.

1906 1907 7.4 Reporting Large Errors 1908 1909 Licensee should correct large errors within 90 days of discovery. The correction may be made by 1910 special submittal or may be submitted with the next (normally scheduled) ARERR (if the next ARERR is 1911 to be submitted within 90 days of discovery of the error). If corrections are made by special submittal, 1912 the licensee should include a brief narrative explaining the errors. The narrative should state that the 1913 affected pages, in their entirety, are included as an attachment. The corrected pages should be attached in 1914 their entirety. The corrected pages should reference the affected calendar year and should contain 1915 revision bars in the margins of the page to indicate the locations of the changes. If submitting corrections 1916 to multiple ARERRs, separate attachment (or addendum) should be made for each of the affected years.

1917 If corrections are made coincident with the next (normally scheduled) submittal of the ARERR, the 1918 correction process should be used as specified in Section 7.2 (for small errors). Other methods of 1919 correcting previous ARERRs may be used provided the corrections are clearly and completely described 1920 consistent with NRC requirements on the completeness and accuracy of information.

1921 RG 1.21, Rev. 3, Page 42

1922 1923 8. Changes to Effluent and Environmental Programs 1924 1925 Standard Technical Specifications (e.g., Section 5.5, Programs and Manuals) establishes 1926 requirements for the radioactive effluent controls and radiological environmental monitoring activities.

1927 The Technical Specifications establish a specific review and approval process for making changes to the 1928 ODCM. Potential changes require licensee analyses or evaluations justifying the change and a 1929 determination that the changes maintain the levels of radioactive effluent control required 1930 by10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I. The evaluation of potential 1931 changes should also consider the need for monitoring in support of decommissioning planning during 1932 operations (see RG 4.22).

1933 1934 Effluent and environmental monitoring programs may need to be modified once power operations 1935 have permanently ceased and a written certification has been submitted to the NRC in accordance with 1936 10 CFR 50.82, Termination of License. The evaluation of potential changes should consider the need 1937 for effluent and environmental monitoring during active decommissioning which is likely to affect 1938 principal release points and principal radionuclides. For example, the removal of effluent ventilation 1939 systems will likely change principal release points and there may be new principal radionuclides 1940 identified (e.g., Kr-85), while radioactive decay may have eliminated former principal radionuclides (e.g.,

1941 I-131) (see Section C.1). Potential changes must be reviewed and approved by the plant manager, station 1942 manager, or as described in plant-specific Technical Specifications, with submittal to the NRC as part of 1943 the next Annual Radioactive Effluent Release Report.

1944 1945 If the plant has a 10 CFR Part 72 ISFSI, the licensee must maintain compliance with the requirements 1946 in 10 CFR Part 72 regarding controls of effluent(s) and an environmental monitoring program. These 1947 requirements include 10 CFR 72.44(d) for 10 CFR Part 72 specific license ISFSIs and, for 1948 10 CFR Part 72 general license ISFSIs, any requirements specified in technical specifications of the 1949 certificate(s) of compliance for the storage systems in use at the ISFSI (to comply with 1950 10 CFR 72.212(b)(3) and (b)(5)).

1951 1952 The radiological criteria for license termination are addressed in 10 CFR 20 Subpart E. The 1953 radiological criteria for unrestricted use (10 CFR 20.1402) encompass contributions from residual 1954 radioactivity in soils and remnant site components and in groundwater. While some reductions in 1955 monitoring programs may be possible when operations cease, other aspects of monitoring such as 1956 groundwater monitoring may need to be increased to adequately characterize residual radioactivity and 1957 characterize dispersion pathways to support dose assessments and to estimate the decommissioning costs.

1958 Lessons learned documented in RG 1.185 and NUREG-1757 indicate that the monitoring data from the 1959 period of operation tend to be insufficient to allow the staff to fully understand the types and the 1960 movement of radioactive material contamination in groundwater at the facility, as well as the extent of the 1961 residual radioactivity. Decommissioning reporting and recordkeeping requirements are addressed in 1962 10 CFR 50.75(g).

1963 1964 Further general guidance to facilitate planning for decommissioning of power plants and 1965 facilities during operations can be found in RG 4.22, in RG 1.185 for post-shutdown decommissioning 1966 activities, in NUREG-1757 for consolidated decommissioning guidance, and in NUREG-1575, Rev. 1, 1967 Multi-Agency Radiation Survey and Site Investigation Manual.

1968 1969 1970 1971 9. Format and Content of the Annual Radioactive Effluent Release Report 1972 RG 1.21, Rev. 3, Page 43

1973 In accordance with 10 CFR 50.4, Written communications, licensees should submit their 1974 annual report electronically or in a written communication. The report should consist of a summary of the 1975 numerical data in a tabular format similar to Tables A-1-A-5 in Appendix A to this RG. Effluent data 1976 reported in Tables A-1, A-1A - A-1F, A-2, A-2A, A-2B, and A-4 should be summarized on a quarterly 1977 and annual basis. Tables A-3 and A-5 should be summarized on an annual basis. In addition to 1978 numerical data, the report should include additional supplemental information containing all the 1979 information in (but not necessarily in the format of Table A-6). Additional detail for the information 1980 contained in each of these tables is listed below. To comply with 10 CFR 50.36a, licensee must submit 1981 their ARERR by May 1 (unless a licensing basis exists for a different submittal date) to report on 1982 effluents and solid waste from the previous calendar year.

1983 1984 Radionuclides that are not detected do not need to be listed in the tables (Tables A-1A-A-1F, A-1985 2A, and A-2B). Activity that is detected should be reported in the appropriate tables (i.e., Tables A-1, A-1986 2, A-1A - A-1F, A-2A, and A-2B) in the ARERR, provided that the amount discharged is numerically 1987 significant with respect to the three-digit exponential format recommended for the ARERR. This should 1988 not be confused with three significant figures. Licensees may round numbers according to accepted 1989 practices (e.g., refer to ASTM E-29, Standard Practice for Using Significant Digits in Test Data to 1990 Determine Conformance with Specifications (Ref. 87)); however, after rounding has been completed, 1991 values should be reported in the ARERR in a three-digit exponential format. Measurements should be 1992 reported for positive values. Some radionuclides that are detected in a year may not be detected in all 1993 quarters. If results are determined to be below detectable levels for an entire quarter, the table entry 1994 should include a suitable designation (e.g., N/D (not detected) and an accompanying footnote) to denote 1995 that measurements were performed but activity was not detected.

1996 1997 The format specified in this RG revision differs slightly from the format specified in Revision 1 1998 and Revision 2. The format and content specified in this Revision 3 of RG 1.21 is one acceptable method 1999 of reporting the data. Other formats may be used (e.g., some tables may be combined) as long as the 2000 specified content is provided (e.g., quarterly totals and annual totals by each release category). However, 2001 licensees are encouraged to use the format listed below to maximize consistency in data reporting. This 2002 format is designed to be consistent with some commonly used electronic-data-reporting software 2003 packages. Consistency in reporting format aids review by members of the public and allows easier 2004 industrywide comparisons of the data.

2005 2006 10 CFR 72 licensees may also, if they choose to do so, use the format specified in this RG for 2007 independent spent fuel storage installation (ISFSI) effluent reports required by 10 CFR 72.44(d) (for 2008 specific licenses) or the storage system(s) certificate(s) of compliance (for general licenses). However, 2009 the ISFSI effluent reporting requirement is not normally satisfied by inclusion as part of the ARERR 2010 since the reporting dates may conflict. If the dates are coincident, or can be met with a single report, 2011 licensees may use the ARERR to fulfill the ISFSI reporting requirements, provided the licensee submits a 2012 copy as specified in those requirements (e.g., 10 CFR 72.44(d)(3) for specific licenses).

2013 2014 9.1 Gaseous Effluents 2015 2016 The quarterly and annual sums of all radionuclides discharged in gaseous effluents (i.e., routine 2017 and abnormal discharges, continuous, and batch) should be reported in a format similar to that of 2018 Tables A-1A - A-1F in Appendix A to this RG. The data should then be further summarized and 2019 reported in the format of Table A-1.

2020 2021 Table A-1, Gaseous EffluentsSummation of All Discharges, contains a summation of all 2022 gaseous effluent discharges from all release points and all modes of release. The data are subdivided by RG 1.21, Rev. 3, Page 44

2023 quarter and year for each radionuclide category: fission and activation gases, iodines/halogens, 2024 particulates, tritium, gross alpha and carbon-14.

2025 Table A-1A, Gaseous EffluentsGround-Level ReleaseBatch Mode, contains a summation 2026 of gaseous effluent releases from ground-level release points in the batch mode of release for six 2027 radionuclide categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha 2028 and carbon-14. Licensees should report the following:

2029 2030 1. curies of each radionuclide discharged by quarter and year, and 2031 2032 2. total curies discharged in each radionuclide category by quarter and year.

2033 2034 Some licensees may have surveillance requirements allowing the non-noble gas radionuclides 2035 (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with 2036 continuous release results. In these instances, the table entries for the affected radionuclides for batch 2037 releases should include an appropriate designation (e.g., *) and an accompanying footnote describing 2038 this situation.

2039 2040 Table A-1B, Gaseous EffluentsGround-Level ReleaseContinuous Mode, contains a 2041 summation of gaseous effluent releases from ground-level release points in the continuous mode of 2042 release for six radionuclide categories: fission and activation gases, iodines/halogens, particulates, 2043 tritium, gross alpha and carbon-14. Licensees should report the following:

2044 2045 1. curies of each radionuclide discharged by quarter and year, and 2046 2. total curies discharged in each radionuclide category by quarter and year.

2047 2048 Table A-1C, Gaseous EffluentsElevated ReleaseBatch Mode, contains a summation of 2049 gaseous effluent releases from elevated release points in the batch mode of release for six radionuclide 2050 categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha, and carbon-2051 14. Licensees should report the following:

2052 2053 1. curies of each radionuclide released by quarter and year, and 2054 2. total curies released in each radionuclide category by quarter and year.

2055 2056 Some licensees may have surveillance requirements allowing the non-noble gas radionuclides 2057 (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with 2058 continuous release results. In these instances, the table entries for the affected radionuclides for batch 2059 releases should include an appropriate designation (e.g., *) and an accompanying footnote describing 2060 this situation.

2061 2062 Table A-1D, Gaseous EffluentsElevated ReleaseContinuous Mode, contains a summation 2063 of gaseous effluent releases from elevated release points in the continuous mode of release for six 2064 radionuclide categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha 2065 and carbon-14. Licensees should report the following:

2066 2067 1. curies of each radionuclide released by quarter and year, and 2068 2069 2. total curies released in each radionuclide category by quarter and year.

2070 2071 Table A-1E, Gaseous EffluentsMixed Mode ReleaseBatch Mode, contains a summation of 2072 gaseous effluent releases from mixed-mode release points in the continuous mode of release for six RG 1.21, Rev. 3, Page 45

2073 radionuclide categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha, 2074 and carbon-14. Licensees should report the following:

2075 2076 1. curies of each radionuclide released by quarter and year, and 2077 2078 2. total curies released in each radionuclide category by quarter and year.

2079 2080 Some licensees may have surveillance requirements allowing the non-noble gas radionuclides 2081 (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with 2082 continuous release results. In these instances, the table entries for the affected radionuclides for batch 2083 releases should include an appropriate designation (e.g., *) and an accompanying footnote describing 2084 this situation.

2085 2086 Table A-1F, Gaseous EffluentsMixed Mode ReleaseContinuous Mode, contains a 2087 summation of gaseous effluent releases from mixed-modes release points in the continuous mode of 2088 release for six radionuclide categories: fission and activation gases, iodines/halogens, particulates, 2089 tritium, gross alpha, and carbon-14. Licensees should report the following:

2090 2091 1. curies of each radionuclide released by quarter and year, and 2092 2093 2. total curies released in each radionuclide category by quarter and year.

2094 2095 9.2 Liquid Effluents 2096 2097 The quarterly and annual sums of all radionuclides discharged in liquid effluents (i.e., routine and 2098 abnormal discharges, continuous, and batch) should be reported in a format similar to that of Tables A-2A 2099 and A-2B. The data should then be further summarized and reported in the format of Appendix A, 2100 Table A-2.

2101 2102 Table A-2, Liquid EffluentsSummation of All Releases, contains a summation of all liquid 2103 radioactive discharges from all release points and all modes of release. The data are subdivided by 2104 quarter and year for each of the radionuclide categories: fission and activation products, tritium, 2105 dissolved and entrained noble gases, and gross alpha.

2106 2107 The table also includes the total volume of primary coolant waste (typically batch mode 2108 releases) before dilution. In this context, primary coolant waste means the higher activity waste that 2109 generally is not discharged directly but is instead typically processed through the liquid radioactive waste 2110 treatment system before discharge. Various methods exist for calculating the dilution water flow rate.

2111 HPPOS-099, Attention to Liquid Dilution Volumes in Semiannual Radioactive Effluent Release 2112 Reports, issued November 1984 (Ref. 88), indicates that licensees should use the total volume of dilution 2113 flow, not just that flow during periods of liquid effluent releases. Licensees should include information 2114 describing how this value is calculated in either the ODCM or the ARERR. Because the primary coolant 2115 waste typically accounts for the vast majority of the radioactivity in liquid waste discharges, the NRC 2116 recommends that the volume and dilution data be summarized separately from the low-activity waste 2117 described in the following paragraph.

2118 2119 The total measured volume or average flow rate of waste from secondary or balance-of-plant 2120 systems (e.g., steam generator blowdown, low-activity waste sumps, and auxiliary boilers) should be 2121 reported. In this context, secondary or balance-of-plant waste means the typically very low-activity waste 2122 that is generally not processed with the liquid radioactive waste treatment system and that collectively 2123 represents a very large volume of waste. Various methods exist for calculating the dilution water flow RG 1.21, Rev. 3, Page 46

2124 rate. HPPOS-099 states that licensees should use the total volume of dilution flow, not just that volume 2125 discharged during periods of liquid effluent releases. Licensees should include information describing 2126 how this value is calculated in either the ODCM or the ARERR. Because of the potentially high volume 2127 and extremely low activity of this type of waste, the NRC recommends the volume and dilution data be 2128 summarized separately from the higher activity waste described in the previous paragraph.

2129 2130 Licensees should report dilution flow rates during periods of release (before effluent is discharged 2131 to the receiving water body), as described above. If calculated differently than described above, the 2132 licensee should describe the method of calculation. Licensees may choose to report near-field dilution if 2133 they account for dilution by the receiving water body. Licensees may report the average, minimum, peak 2134 river, and stream flow rates, as applicable.

2135 2136 Table A-2A, Liquid EffluentsBatch Mode, contains a summation of liquid effluent 2137 discharges in the batch mode of release. The table is divided into four radionuclide categories: fission 2138 and activation products, tritium, dissolved and entrained gases, and gross alpha. Licensees should report 2139 the following:

2140 2141 1. curies of each radionuclide and gross alpha discharged by quarter and year, and 2142 2143 2. total curies in each radionuclide category by quarter and year.

2144 2145 Table A-2B, Liquid EffluentsContinuous Mode, contains a summation of liquid effluent 2146 discharges in the continuous mode of release. The table is divided into four radionuclide categories:

2147 fission and activation products, tritium, dissolved and entrained gases, and gross alpha. Licensees should 2148 report the following:

2149 2150 1. curies of each radionuclide and gross alpha discharged by quarter and year, and 2151 2152 2. total curies in each radionuclide category by quarter and year.

2153 2154 9.3 Solid Waste Shipments Released from the Unit (per Standard Technical Specifications) 2155 2156 Appendix A, Table A-3, provides an acceptable format for reporting the solid radioactive waste 2157 released (shipped) from the unit (plant site) during the reporting period. The NRC intends that licensees 2158 report the waste shipped from the site, regardless of whether the shipment is sent for waste processing or 2159 direct disposal (i.e., with or without waste processing).

2160 2161 Licensee should report the volume and curies of solid waste shipped (see exceptions noted in 2162 Section 6) for each of the following waste streams:

2163 2164 1. wet radioactive waste (e.g., spent resins, filters, sludges, etc.),

2165 2166 2. dry radioactive waste (e.g., trash, paper, discarded protective clothing, etc.),

2167 2168 3. activated or contaminated radioactive material (e.g., equipment or bulk radioactive material, etc.),

2169 and 2170 2171 4. other radioactive waste (waste not included in the above categories and waste not excepted from 2172 reporting requirements in Section 6).

2173 RG 1.21, Rev. 3, Page 47

2174 9.4 Dose Assessments 2175 2176 Licensees should calculate the annual evaluations of dose to members of the public using 2177 RG 1.21, Section 5 and report the data in the format of Tables A-4 and A-5. Dose assessments should 2178 demonstrate compliance with the following9:

2179 2180 1. Licensees should demonstrate compliance with 10 CFR Part 50, Appendix I (see Table A-4), by 2181 doing the following10:

2182 2183 a. Reporting the calculated dose from liquid effluents on a quarterly and annual basis to the 2184 total body and maximum organ and the percentage of the 10 CFR Part 50, Appendix I, 2185 design objectives for the maximum exposed individual. If a particular exposure pathway 2186 is not applicable (i.e., it does not exist at a site), do not calculate the dose for that 2187 exposure pathway.

2188 2189 b. Reporting the highest air dose from gaseous effluents on a quarterly and annual basis at 2190 any location that could be occupied by individuals in the unrestricted area and the 2191 percentage of the 10 CFR Part 50, Appendix I, design objectives.

2192 2193 c. Reporting the organ dose from iodine, tritium, and particulates with a half-life greater 2194 than 8 days to the maximum exposed individual in an unrestricted area from all pathways 2195 of exposure (e.g., submersion and ingestion).

2196 2197 2. Licensees must demonstrate compliance with 10 CFR 20.1301(e) and 40 CFR Part 190 (see 2198 Table A-5) as follows:

2199 2200 a. Reporting the whole body, thyroid, and highest dose to any other organ from licensed and 2201 unlicensed radioactive material in the uranium fuel cycle, excluding background, to the 2202 individual member of the public likely to receive the highest dose.

2203 2204 9.5 Supplemental Information 2205 2206 Licensees should provide supplemental information in a descriptive, narrative form (see 2207 Table A-6 or in a similar format). Relevant information and a description of circumstances should be 2208 provided as appropriate for each the following categories, adding categories as appropriate. The 2209 annotation N/A should be used if a category is not applicable.

2210 2211 9.5.1 Abnormal Releases or Abnormal Discharges 2212 2213 The reporting of abnormal releases to onsite areas and abnormal discharges to unrestricted areas 2214 should include the following:

2215 2216 1. Specific information should be reported concerning abnormal (airborne, liquid) releases on site 2217 and abnormal discharges to the unrestricted area. The report should describe each event in a way 2218 that would enable the NRC to adequately understand how the material was released and if there 2219 was a discharge to the unrestricted area. The report should describe the potential impact on the 9 As noted in Section C.5, dose assessments for 10 CFR 72.104 should include the components necessary to appropriately demonstrate compliance with those limits.

10 The type of individual or dose receptor should be identified as a real individual or as a hypothetical individual if using bounding dose assessments; the individual/ receptor is in the unrestricted area.

RG 1.21, Rev. 3, Page 48

2220 ingestion exposure pathway involving surface water and groundwater, as applicable. The report 2221 should also describe the impact (if any) on other affected exposure pathways (e.g., inhalation).

2222 2223 2. The following are the thresholds for reporting abnormal releases and abnormal discharges in the 2224 supplemental information section:

2225 2226 a. abnormal releases or abnormal discharges that are voluntarily reported to local authorities 2227 under Nuclear Energy Institute 07-07, Revision 1, Industry Groundwater Protection 2228 InitiativeFinal Guidance Document, dated February 26, 2019 (Ref. 89);

2229 2230 b. abnormal releases or abnormal discharges estimated to exceed 300 liters (100 gallons) of 2231 radioactive liquid where the presence of licensed radioactive material is positively 2232 identified (in either the onsite environs or in the source of the leak or spill) as greater than 2233 the minimum detectable activity11 for the laboratory instrumentation; 2234 2235 c. abnormal releases to onsite areas that result in detectable residual radioactivity after 2236 remediation; 2237 2238 d. abnormal releases that result in a high effluent radiation alarm without an anticipated 2239 system trip occurring; and 2240 2241 e. abnormal discharges to an unrestricted area.

2242 2243 3. Information on abnormal releases or abnormal discharges should include the following, as 2244 applicable:

2245 2246 a. date and duration, 2247 b. location, 2248 c. volume, 2249 d. estimated activity of each radionuclide, 2250 e. effluent monitoring results (if any),

2251 f. onsite monitoring results (if any),

2252 g. depth to the local water table, 2253 h. classification(s) of subsurface aquifer(s) (e.g., drinking water, unfit for drinking water, 2254 not used for drinking water),

2255 i. size and extent of any groundwater plume, 2256 j. expected movement/mobility of any groundwater plume, 2257 k. land use characteristics (e.g., water used for irrigation),

2258 l. remedial actions considered or taken and results obtained, 2259 m. calculated member of the public dose attributable to the release, 2260 n. calculated member of the public dose attributable to the discharge, 2261 o. actions taken to prevent recurrence, as applicable, and 2262 p. whether the NRC was notified, the date(s), and the contact organization.

2263 2264 9.5.2 Nonroutine Planned Discharges 2265 11 The minimum detectable activity is a post-analysis calculation of sensitivity level based on the actual sample measurement.

RG 1.21, Rev. 3, Page 49

2266 Discharges resulting from remediation efforts that are not identified in the ODCM should be 2267 reported. For example, the remediation effort may include pumping of contaminated groundwater in 2268 response to leaks and spills.

2269 2270 9.5.3 Radioactive Waste Treatment System Changes 2271 2272 Licensees should report any changes or modifications affecting any portion of the gaseous 2273 radioactive waste treatment system, the ventilation exhaust treatment system, or the liquid radioactive 2274 waste treatment.

2275 2276 9.5.4 Annual Land Use Census Changes 2277 2278 Licensees should report any changes or modifications affecting significant aspects of the 2279 environmental monitoring program such as receptors, receptor locations, sample media availability, or 2280 new (or changed) routes of exposure.

2281 2282 9.5.5 Effluent Monitoring System Inoperability 2283 2284 Licensees should report information on inoperable effluent monitors as follow:

2285 2286 1. If an effluent radiation monitor is not operable for the consecutive time period listed in 2287 the licensees ODCM or technical specifications (typically 30 days), then the ARERR 2288 should include the radiation monitors equipment designation, the common name of the 2289 effluent radiation monitor, the time period of the inoperability, the reason why this 2290 inoperability was not corrected in a timely manner, and any other information required by 2291 the licensees ODCM or technical specifications.

2292 2293 2. In accordance with NUREG-1301 and NUREG-1302, Sections 3.3.3.10.b and 3.3.3.11.b, 2294 Generic Letter 89-01, and licensee ODCMs, the information above is required only when 2295 the minimum channels operability requirement is not achieved for the consecutive time 2296 period listed in the ODCM (typically 30 days).

2297 2298 9.5.6 Offsite Dose Calculation Manual Changes 2299 2300 Licensees should report any changes or modifications affecting significant aspects of the ODCM.

2301 2302 9.5.7 Process Control Program Changes 2303 2304 Licensees should report any changes or modifications affecting significant aspects of the ODCM.

2305 2306 9.5.8 Corrections to Previous Reports 2307 2308 When submitting corrections to previous reports, licensees should do the following:

2309 2310 1. Include a brief explanation of the error(s).

2311 2312 2. State that the affected pages, in their entirety, are included as attachments to this ARERR.

2313 2314 3. Ensure that a copy of the affected page(s), in their entirety, is included as an attachment 2315 to the ARERR. The attached pages should reference the affected calendar year and 2316 contain revision bars.

RG 1.21, Rev. 3, Page 50

2317 2318 9.5.9 Other (Narrative Descriptions of Other Information Related to Radioactive Effluents) 2319 2320 Licensees should report other supplemental information (as appropriate).

RG 1.21, Rev. 3, Page 51

2321 D. IMPLEMENTATION 2322 2323 The NRC staff may use this regulatory guide as a reference in its regulatory processes, such as 2324 licensing, inspection, or enforcement. However, the NRC staff does not intend to use the guidance in this 2325 regulatory guide to support NRC staff actions in a manner that would constitute backfitting as that term is 2326 defined in 10 CFR 50.109, Backfitting, or in 10 CFR 72.62, Backfitting, and as described in NRC 2327 Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information 2328 Requests, (Ref. 90), nor does the NRC staff intend to use the guidance to affect the issue finality of an 2329 approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The 2330 staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes 2331 forward fitting as that term is defined and described in Management Directive 8.4. If a licensee believes 2332 that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this 2333 Implementation section, then the licensee may file a backfitting or forward fitting appeal with the NRC in 2334 accordance with the process in Management Directive 8.4.

2335 2336 2337 RG 1.21, Rev. 3, Page 52

2338 GLOSSARY 2339 2340 a prioriBefore-the-fact limit, representing the capability of a measurement system and not as an 2341 after-the-fact (a posteriori) limit for a particular measurement.

2342 2343 abnormal dischargeThe unplanned or uncontrolled emission of an effluent (i.e., containing 2344 plant-related, licensed radioactive material) into the unrestricted area.

2345 2346 abnormal releaseThe unplanned or uncontrolled emission of an effluent (i.e., containing plant-related, 2347 licensed radioactive material) into the onsite environs.

2348 2349 accumulated radioactivityRadioactivity from prior-year effluent releases that may still be present in 2350 the media of concern.

2351 2352 background (radiation)Means radiation from cosmic sources; naturally occurring radioactive 2353 material, including radon (except as a decay product of source or special nuclear material); and 2354 global fallout as it exists in the environment from the testing of nuclear explosive devices and 2355 from past nuclear accidents, such as Chernobyl, that contribute to background radiation and are 2356 not under the control of the licensee. Background radiation does not include radiation from 2357 source, byproduct, or special nuclear materials regulated by the Commission.

2358 2359 batch releaseThe release of liquid (radioactive) wastes of a discrete volume or the release of a tank or 2360 purge of radioactive gases into the site environs.

2361 2362 channel checkThe qualitative assessment of channel behavior during operation by observation. This 2363 determination should include, where possible, comparison of the channel indication, status with 2364 other indications, and status derived from independent instrument channels measuring the same 2365 parameter.

2366 2367 channel operational testThe injection of a simulated signal into the channel as close to the sensor as 2368 practicable to verify operability of alarm, interlock, and trip functions, as applicable. The channel 2369 operational test should include adjustments, as necessary, of the alarm, interlock, and trip 2370 setpoints, as applicable, such that the setpoints are within the required range and accuracy.

2371 2372 continuous releaseAn essentially uninterrupted release of gaseous or liquid effluent for extended 2373 periods during normal operation of the facility where the volume of radioactive waste is 2374 non-discrete and there is input flow during the release.

2375 2376 controlled area (10 CFR Part 20)An area outside of a restricted area but inside the site boundary, 2377 access to which is limited by the licensee for any reason.

2378 2379 controlled area (10 CFR Part 72)The area immediately surrounding an ISFSI or a monitored 2380 retrievable storage installation (MRS) for which the licensee exercises authority over its use and 2381 within which ISFSI or MRS operations are performed.

2382 2383 controlled dischargeA radioactive discharge is considered to be controlled if (1) the discharge was 2384 conducted in accordance with methods, and without exceeding any of the limits, outlined in the 2385 ODCM or (2) if one or more of the following three items are true:

2386 RG 1.21, Rev. 3, Page 53

2387 1. The radioactive discharge had an associated, preplanned method of radioactivity 2388 monitoring that assured the discharge was properly accounted and was within the limits 2389 set by 10 CFR Part 20 and 10 CFR Part 50.

2390 2391 2. The radioactive discharge had an associated, preplanned method of termination (and 2392 associated termination criteria) that assured the discharge was properly accounted and 2393 was within the limits set by 10 CFR Part 20 and 10 CFR Part 50.

2394 2395 3. The radioactive discharge had an associated, preplanned method of adjusting, 2396 modulating, or altering the flow rate (or the rate of release of radioactive material) that 2397 assured the discharge was properly accounted and was within the limits set by 2398 10 CFR Part 20 and 10 CFR Part 50.

2399 2400 controlled releaseA radioactive release is considered to be controlled if (1) the release was 2401 conducted in accordance with methods, and without exceeding any of the limits, outlined in the 2402 ODCM or (2) if one or more of the following three items are true:

2403 2404 1. The radioactive release had an associated, preplanned method of radioactivity monitoring 2405 that assured the release was properly accounted and was within the limits set by 2406 10 CFR Part 20 and 10 CFR 50.

2407 2408 2. The radioactive release has an associated, preplanned method of termination 2409 (and associated termination criteria) that assured the release was properly accounted and 2410 was within the limits set by 10 CFR Part 20 and 10 CFR 50.

2411 2412 3. The radioactive release had an associated, preplanned method of adjusting, modulating, 2413 or altering the flow rate (or the rate of release of radioactive material) that assured the 2414 release was properly accounted and was within the limits set by10 CFR Part 20 and 2415 10 CFR Part 50.

2416 2417 conversion factorA factor (e.g., microcuries per cubic centimeter per counts per minute (Ci/cc/cpm) 2418 used to estimate a radioactivity concentration in an effluent based on a gross radioactivity 2419 measurement (e.g., cpm).

2420 2421 D/QA deposition value used for estimating the dose to an individual at a specified (e.g., controlling) 2422 location. D/Q may be described as the downwind surface or ground deposition (D) (e.g., in units 2423 of microcuries per square meter [Ci/m2]) of radioactive material at a location, divided by the 2424 release activity (Q) (e.g., in Ci). D/Q is thus a normalized downwind surface deposition value 2425 per unit release and can be used to determine the surface or ground radioactivity concentration 2426 during a measured effluent release over a specific period of time. The units of D/Q are reciprocal 2427 square meters.

2428 2429 determinationA quantitative evaluation of the release or presence of radioactive material under a 2430 specific set of conditions. A determination may be made by direct measurement or indirect 2431 measurements (e.g., with the use of scaling factors).

2432 2433 dilution water (for liquid radioactive waste)For purposes of this RG, any water other than the 2434 undiluted radioactive waste that is mixed with undiluted liquid radioactive waste before its 2435 ultimate discharge to the unrestricted area.

2436 RG 1.21, Rev. 3, Page 54

2437 discharge pointA location at which radioactive material enters the unrestricted area. This would be 2438 the point beyond the vertical plane of the unrestricted area (surface or subsurface).

2439 2440 drinking waterWater that does not contain an objectionable pollutant, contamination, minerals, or 2441 infective agent and is considered satisfactory for domestic consumption. This is sometimes called 2442 potable water. Potable water is water that is safe and satisfactory for drinking and cooking.

2443 Although EPA regulations only apply to public drinking water sources supplying 25 or more 2444 people (refer to the EPA for more information), for purposes of the effluent and environmental 2445 monitoring programs, the term drinking water includes water from single-use residential drinking 2446 water wells.

2447 2448 effluentLiquid or gaseous waste containing plant-related, licensed radioactive material, emitted at the 2449 boundary of the facility (e.g., buildings, end-of-pipe, stack, or container) as described in the final 2450 safety analysis report.

2451 2452 effluent dischargeThe portion of an effluent release that reaches an unrestricted area. (See also the 2453 definition for radioactive discharge.)

2454 2455 effluent releaseThe emission of an effluent. (See also the definition for radioactive release.)

2456 2457 elevated releaseA gaseous effluent release made from a height that is more than twice the height of 2458 adjacent solid structures, or releases made from heights sufficiently above adjacent solid 2459 structures such that building wake effects are minimal or absent.

2460 2461 exposure pathwayA mechanism by which radioactive material is transferred from the (local) 2462 environment to humans. There are three commonly recognized exposure pathways: inhalation, 2463 ingestion, and direct radiation. For example, ingestion may include dose contributions from one 2464 or more routes of exposure. One route of exposure that may contribute to the ingestion exposure 2465 pathway is often referred to as grass-cow-milk-infant-thyroid route of exposure.

2466 2467 general environmentAn EPA 40 CFR 190.02 definition meaning the total terrestrial, atmospheric and 2468 aquatic environment outside sites upon which any (licensed) operation of a nuclear fuel cycle is 2469 conducted.

2470 2471 ground-level releaseA gaseous effluent release made from a height that is ator less thanthe height 2472 of adjacent solid structures, or where the degree of plume rise is unknown or is otherwise 2473 insufficient to avoid building wake effects.

2474 2475 groundwaterAll water in the surface soil, the subsurface soil, or any other subsurface water.

2476 Groundwater is simply water in the ground regardless of its quality, including saline, brackish, or 2477 fresh water. Groundwater can be moisture in the ground that is above the regional water table in 2478 the unsaturated (or vadose) zone, or groundwater can be at and below the water table in the 2479 saturated zone.

2480 2481 hypothetical exposure pathwayAn exposure pathway in which one or more of the components 2482 involved in the transfer of a radionuclide from the environment to the human does not actually 2483 exist at the specified location, or if a real human does not consume, inhale, or otherwise become 2484 exposed to the radioactive material. For example, the grass-cow-milk-infant-thyroid route of 2485 exposure (associated with the ingestion exposure pathway) would be considered a hypothetical 2486 exposure pathway if the grass, the cow, or the milk did not actually exist at a specified location or 2487 if an infant did not actually consume the milk.

RG 1.21, Rev. 3, Page 55

2488 2489 impacted areasThe areas with some reasonable potential for residual radioactivity in excess of natural 2490 background or fallout levels. The NRC discusses impacted areas in 10 CFR 50.2 and 2491 NUREG-1757. For example, impacted areas include locations where radiological leaks or spills 2492 have occurred within the onsite environs (i.e., outside of the facilitys systems, structures, and 2493 components). (See also the definition for significant contamination.)

2494 2495 leachateWater containing contaminants that is percolating downward from a pond or lake into the 2496 subsurface.

2497 2498 less-significant release pointAny location from which radioactive material is released as a liquid or 2499 gaseous effluent contributing less than or equal to 1 percent of the activity discharged from all the 2500 release points for a particular type of effluent considered. RG 1.109 lists the three types of 2501 effluent as (1) liquid effluents, (2) noble gases discharged to the atmosphere in gaseous 2502 radioactive waste, and (3) all other nuclides discharged to the atmosphere in gaseous radioactive 2503 waste.

2504 2505 Example: If 1,000 curies (Ci) of tritium are released in all liquid effluents in a given period of 2506 time (e.g., a typical calendar year or fuel cycle) and 0.01 Ci of tritium is released in steam 2507 generator blowdown, then the steam generator blowdown would be a less-significant release 2508 point. Similarly, for gaseous releases of radionuclides other than noble gases (i.e., iodine, 2509 particulates, and tritium), if the total effluents are 10 Ci (iodine, particulates, and tritium), and the 2510 refueling water storage tank released 0.009 Ci of iodine, particulates, and tritium, then the 2511 refueling water storage tank would be a less-significant release point. In both examples, the 2512 sample frequency can be adjusted to an appropriate frequency for the less-significant release 2513 point. Samples collected from these systems for other programs (e.g., detection of 2514 primary-to-secondary leakage) must still be collected and analyzed at the frequencies specified by 2515 the other programs.

2516 2517 licensed materialSource material, special nuclear material, or byproduct material received, possessed, 2518 used, transferred, or disposed under a general or specific license issued by the Commission.

2519 2520 lower limit of detection (LLD)The a priori smallest concentration of radioactive material in a sample 2521 that will yield a net count, above system background, that will be detected with 95-percent 2522 probability with only a 5-percent probability of falsely concluding that a blank observation 2523 represents a real signal (see NUREG-1301, NUREG-1302, and NUREG/CR-4007, Lower Limit 2524 of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and 2525 Environmental Measurements, issued September 1984 (Ref. 91)).

2526 2527 maximum exposed individualIndividuals characterized as maximum exposed with regard to food 2528 consumption, occupancy, and other usage in the vicinity of the plant site. As such, the maximum 2529 exposed individual represents individuals with habits that are considered to be maximum 2530 reasonable deviations from the average for the population in general. Additionally, in 2531 physiological or metabolic respects, the maximum exposed individual is assumed to have those 2532 characteristics that represent the averages for the corresponding age group in the general 2533 population. (This term typically refers to members of the public.) RG 1.109 contains additional 2534 information.

2535 2536 member of the public (10 CFR Part 20)Any individual except when that individual is receiving an 2537 occupational dose.

2538 RG 1.21, Rev. 3, Page 56

2539 member of the public (40 CFR Part 190)Any individual that can receive a radiation dose in the 2540 general environment, whether the individual may or may not also be exposed to radiation in an 2541 occupation associated with a nuclear fuel cycle. However, an individual is not considered a 2542 member of the public during any period in which the individual is engaged in carrying out any 2543 operation that is part of a nuclear fuel cycle.

2544 2545 minimum detectable concentrationThe smallest activity concentration measurement that is 2546 practically achievable with a given instrument and type of measurement procedure. The 2547 minimum detectable concentration depends on factors involved in the survey measurement 2548 process (surface type, geometry, backscatter, and self-absorption) and is typically calculated 2549 following an actual sample analysis (a posteriori). (See NUREG-1507, Minimum Detectable 2550 Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field 2551 Conditions, issued June 1998 (Ref. 92)).

2552 2553 mixed mode releaseA gaseous effluent release made from a height higher than a ground-level release 2554 but less than an elevated release where, sometimes, because of a lack of plume rise 2555 (e.g., buoyancy, momentum, wind speed), a proper estimate of radionuclide transport and 2556 diffusion requires mathematically splitting the plume into (1) an elevated component and (2) a 2557 ground-level component to properly account for building wake effects, release, or ambient 2558 conditions (or a combination of all three). (RG 1.111 contains further guidance.)

2559 2560 monitoringWith respect to radiation or radiation protection, the measurement of radiation levels, 2561 concentrations, surface area concentrations, or quantities of radioactive material and the use of 2562 results of these measurements to evaluate potential exposures and doses.

2563 2564 nonroutine, planned dischargeAn effluent release from a release point not defined in the ODCM but 2565 that has been planned, monitored, and discharged in accordance with 10 CFR 20.2001 (e.g., the 2566 discharge of water recovered during a spill or leak from a temporary storage tank).

2567 2568 nuclear fuel cycleThe operations defined to be associated with the production of electrical power for 2569 public use by any fuel cycle through the use of nuclear energy (see 40 CFR 190.02).

2570 2571 onsite environsLocation within the site boundary but outside of the systems, structures, or components 2572 described in the final safety analysis report or the ODCM.

2573 2574 operability (operable)The ability of a system, subsystem, train, component, or device to perform its 2575 specified safety function(s) and the ability of all necessary attendant instrumentation, controls, 2576 normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary 2577 equipment (required for the system, subsystem, train, component, or device to perform its 2578 specified safety function(s)) to perform their related support function(s).

2579 2580 principal radionuclideOne of the principal gamma emitters listed in NUREG-1301 and 2581 NUREG-1302, Tables 4.11-1 and 4.11-2, or, from a risk-informed perspective, a radionuclide that 2582 contributes either (1) greater than 1 percent of the 10 CFR Part 50, Appendix I, design objective 2583 dose when all radionuclides in the type of effluent are considered, or (2) greater than 1 percent of 2584 the activity of all nuclides in the type of effluent being considered. RG 1.109 lists the three types 2585 of effluents as (1) liquid effluents, (2) noble gases discharged to the atmosphere, and (3) all other 2586 nuclides discharged to the atmosphere. In this RG, the terms principal radionuclide and 2587 principal nuclide are synonymous since this document is only concerned with measuring, 2588 evaluating, and reporting radioactive materials in effluents.

2589 RG 1.21, Rev. 3, Page 57

2590 radioactive dischargeThe emission of an effluent (i.e., containing plant-related, licensed radioactive 2591 material) into the unrestricted area. (See also the definition for effluent discharge.)

2592 2593 radioactive releaseThe emission of an effluent (i.e., containing plant-related, licensed radioactive 2594 material). (See also the definition for effluent release.)

2595 2596 real exposure pathwayAn exposure pathway in which plant-related radionuclides in the environment 2597 at (or from) a specified location cause exposure to an actual individual. For example, the 2598 grass-cow-milk-infant-thyroid exposure pathway would be considered a real exposure pathway if 2599 the grass, the cow, and the milk actually existed at a specified location and an infant actually 2600 consumed the milk. For purposes of compliance with 10 CFR Part 50, Appendix I, the individual 2601 must be a member of the public.

2602 2603 real individual (10 CFR 72) - Any individual who lives, works, or engages in recreation or other 2604 activities close to the ISFSI/MRS for a significant portion of the year.

2605 2606 release sourceA system, structure, or component (containing radioactive material under the licensees 2607 control) where radioactive materials are contained before release.

2608 2609 release pointA location from which radioactive materials are released from a system, structure, or 2610 component (including evaporative releases and leaching from ponds and lakes in the controlled or 2611 restricted area before release under 10 CFR 20.2001). For release points monitored by plant 2612 process radiation monitoring systems, the release point is associated with the piping immediately 2613 downstream of the radiation monitor. (See also the definition for significant release point.)

2614 Several release sources may contribute to a common release point.

2615 2616 residual radioactivityRadioactivity in structures, materials, soils, groundwater, and other media at a 2617 site resulting from activities under the licensees control. This includes radioactivity from all 2618 licensed and unlicensed sources used by the licensee, but it excludes background radiation. It 2619 also includes radioactive materials remaining at the site as a result of routine or accidental 2620 releases of radioactive material at the site and previous burials at the site, even if those burials 2621 were made in accordance with 10 CFR Part 20.

2622 2623 restricted areaAn area, access to which is limited by the licensee for the purpose of protecting 2624 individuals against undue risks from exposure to radiation and radioactive materials. Restricted 2625 area does not include areas used as residential quarters, but separate rooms in a residential 2626 building may be set apart as a restricted area.

2627 2628 route of exposureA specific path (or delivery mechanism) by which radioactive material can 2629 eventually cause a radiation dose to an individual. The path typically includes a type of 2630 environmental medium (e.g., air, grass, meat, or water) as the starting point and a recipients 2631 organ or body as the end point. For example, the grass-cow-milk-infant-thyroid route of exposure 2632 may contribute to the ingestion exposure pathway. Additionally, several routes of exposure may 2633 contribute to a single exposure pathway.

2634 2635 scaling factorA factor used to estimate the unknown activity of a radionuclide based on its ratio to the 2636 activity of a readily measured radionuclide or other parameter (e.g., carbon-14 scaled to power 2637 generation).

2638 2639 significant contaminationAs used for 10 CFR 50.75(g) recordkeeping, a quantity, concentration, or 2640 both, of residual radioactivity that would require remediation during decommissioning in order to RG 1.21, Rev. 3, Page 58

2641 terminate the license by meeting the unrestricted use criteria stated in 10 CFR 20.1402 (see 2642 NUREG-1757).

2643 2644 significant release pointAny location from which radioactive material is released that contributes 2645 greater than 1 percent of the activity discharged from all the release points for a particular type of 2646 effluent considered. RG 1.109 lists the three types of effluent as (1) liquid effluents, (2) noble 2647 gases discharged to the atmosphere in gaseous radioactive waste, and (3) all other radionuclides 2648 discharged to the atmosphere in gaseous radioactive waste.

2649 2650 significant residual radioactivitySee the definition for significant contamination.

2651 2652 site boundaryThe line beyond which the licensee owns, leases, or otherwise controls land or property.

2653 2654 site environsLocations outside of the nuclear power plant systems, structures, or components as 2655 described in the final safety analysis report or the ODCM.

2656 2657 solid radioactive waste (solid waste)solid material for which the licensee foresees no further use.

2658 2659 source checkA qualitative assessment of the channel response when the channel sensor is exposed to a 2660 source of increased radioactivity.

2661 2662 standard (instrument or source) (see ANSI N323C-2009 and ANSI N42.22-2006, Traceability of 2663 Radioactive Sources to the National Institute of Standards and Technology (NIST) and 2664 Associated Instrument Quality Control (Ref. 93):

2665 2666

• National standarda standard determined by a nationally recognized, competent 2667 authority to serve as the basis for assigning values to other standards of the quantity 2668 concerned. In the United States, this is an instrument, source, or other system or device 2669 maintained and promulgated by the NIST.

2670 2671

• Primary standarda standard that is designated or widely acknowledged as having the 2672 highest metrological qualities and whose value is accepted without reference to other 2673 standards of the same quantity.

2674 2675

• Secondary standarda standard whose value is assigned by comparison with a primary 2676 standard of the same quantity.

2677 2678

• Reference standarda standard, generally having the highest metrological quality 2679 available at a given location or in a given organization, from which measurements made 2680 there are derived.

2681 2682

• Transfer standardA standard used as an intermediary to compare standards. (If the 2683 intermediary is not a standard, the term transfer device should be used.)

2684 2685

• Working standarda standard that is used routinely to calibrate or check material 2686 measures, measuring instruments, or reference materials. A working standard is usually 2687 traceable to the NIST.

2688 2689 surveyAn evaluation of the radiological conditions and potential hazards incident to the production, 2690 use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

RG 1.21, Rev. 3, Page 59

2691 When appropriate, such an evaluation includes a physical survey of the location of radioactive 2692 material and measurements or calculations of levels of radiation, or concentrations or quantities 2693 of radioactive material present.

2694 2695 type of effluentA grouping of radioactive releases into one of the three categories listed in 2696 10 CFR Part 50, Appendix I, paragraphs A-C. RG 1.109 classifies the three categories as 2697 (1) liquid effluents, (2) noble gases discharged to the atmosphere in gaseous radioactive waste, 2698 and (3) all other nuclides discharged to the atmosphere in gaseous radioactive waste.

2699 2700 unlicensed materialRadioactive material discharged as licensed material in effluents and background 2701 radioactivity (including global fallout). Licensed radioactive material becomes unlicensed 2702 radioactive material upon discharge in effluents, in accordance with 10 CFR 20.2001.

2703 2704 uncontrolled dischargeAn effluent discharge that does not meet the definition of a controlled 2705 discharge. (See also the definition of controlled discharge.)

2706 2707 uncontrolled releaseAn effluent release that does not meet the definition of a controlled release. (See 2708 also the definition of controlled release).

2709 2710 unplanned dischargeThe unintended or unexpected discharge of liquid or airborne radioactive 2711 material to the unrestricted area. Examples of an unplanned discharge include the following:

2712 2713

• the unintentional discharge of a wrong waste gas decay tank (or bulk liquid radioactive 2714 waste tank), or 2715 2716

• the failure of a radiation monitor to divert liquid to the radioactive waste system in the 2717 case where radioactivity is present and the automatic alarm/trip function fails to divert 2718 material to liquid radioactive waste and that material (or a portion of that material) 2719 instead discharges to the environment.

2720 2721 unplanned releaseThe unintended or unexpected release of liquid or airborne radioactive material to 2722 the onsite environment. An example of an unplanned release would include a plant occurrence 2723 that results in a leak or spill of radioactive material to onsite areas, requiring a report under 2724 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors, or 2725 10 CFR 50.73, License event report system. (See HPPOS-254, Definition of Unplanned 2726 Release, issued February 1994 (Ref. 94)).

2727 2728 For example, if a licensee has prepared documents describing an intended release (e.g., a 2729 preliminary radioactive waste release permit) in advance of the evolution, and the intended 2730 release occurs as planned, then the release is a planned release. If such documents (e.g., a 2731 preliminary release permit) are not prepared (or considered/evaluated) before the release, it is 2732 potentially an unplanned release (and additional information may be required to determine if it is 2733 an unplanned release).

2734 2735 unrestricted areaAn area for which the licensee neither limits nor controls access.

2736 2737 uranium fuel cycleThe operations of milling of uranium ore, chemical conversion of uranium, isotopic 2738 enrichment of uranium, fabrication of uranium fuel, generation of electricity by a 2739 light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium 2740 fuel, to the extent that these directly support the production of electrical power for public use 2741 using nuclear energy, but excludes mining operations, operations at waste disposal sites, RG 1.21, Rev. 3, Page 60

2742 transportation of any radioactive material in support of these operations, and the reuse of 2743 recovered nonuranium special nuclear and byproduct materials from the cycle.

2744 2745 /QReferred to as Chi over Q, the average atmospheric effluent concentration, , normalized by 2746 release rate, Q, at a distance (or location) in a given downwind direction. Expressed in another 2747 way, /Q is the concentration () of airborne radioactive material (e.g., in units of Ci/m3) divided 2748 by the release rate (Q) (e.g., in units of Ci/s) at a specified distance and direction downwind of 2749 the release point.

RG 1.21, Rev. 3, Page 61

2750 REFERENCES12 2751 2752 1. U.S. Code of Federal Regulations (CFR), Standards for Protection Against Radiation, Part 20, 2753 Chapter 1, Title 10, Energy.

2754 2755 2. CFR, Environmental Radiation Protection Standards for Nuclear Power Operations, Part 190, 2756 Chapter 1, Title 40, Protection of Environment.

2757 2758 3. U.S. Nuclear Regulatory Commission (NRC), Staff RequirementsSECY-98-144White 2759 Paper on Risk Informed and Performance-Based Regulation, SRM-SECY-98-144, 2760 February 24, 1999 ADAMS Accession No. ML003753593.

2761 2762 4. CFR, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, 2763 Energy.

2764 2765 5. CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter 1, 2766 Title 10, Energy.

2767 2768 6. CFR, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level 2769 Radioactive Waste, and Reactor-Related Greater Than Class C Waste, Part 72, Chapter 1, 2770 Title 10, Energy.

2771 2772 7. CFR, Environmental Radiation Protection Standards for Management and Disposal of Spent 2773 Nuclear Fuel and Transuranic Radioactive Wastes, Part 191, Chapter 1, Title 40, Standards.

2774 2775 8. NRC, Regulatory Guide (RG) 1.23, Meteorological Monitoring Programs for Nuclear Power 2776 Plants, Revision 1, March 2007.

2777 2778 9. NRC, RG 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant 2779 and Environs Conditions During and Following an Accident, Revision 0, December 1975; 2780 Revision 1, August 1977; Revision 2, December 1980; and Revision 3, May 1983.

2781 2782 10. NRC, RG 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, 2783 Regulatory Guide 1.97, Revision 4, June 2006.

2784 2785 11. IEEE, Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating 2786 Stations, Std. 497-2002, New York, NY.

2787 2788 12. NRC, RG 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, 2789 Regulatory Guide 1.97, Revision 5, April 2019.

2790 2791 13. IEEE, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power 2792 Generating Stations, Std. 497-2016, New York, NY.

2793 2794 14. NRC, RG 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor 12 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/index.html and through the NRCs ADAMS at http://www.nrc.gov/reading-rm/adams.html. The documents can also be viewed online or printed for a fee in the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or (800) 397-4209; fax (301) 415-3548; or e-mail pdr.resource@nrc.gov.

RG 1.21, Rev. 3, Page 62

2795 Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, 2796 Revision 1, October 1977.

2797 2798 15. NRC, RG 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous 2799 Effluents in Routine Releases from Light-Water-Cooled Reactors, Revision 1, July 1977.

2800 2801 16. NRC, RG 1.112, Calculation of Releases of Radioactive Materials in Gaseous and Liquid 2802 Effluents from Light-Water-Cooled Power Reactors, Revision 1, March 2007.

2803 2804 17. NRC, RG 1.113, Estimating Aquatic Dispersion of Effluents from Accidental and Routine 2805 Reactor Releases for the Purpose of Implementing Appendix I, Revision 1, April 1977.

2806 2807 18. NRC, RG 1.184, Decommissioning of Nuclear Power Reactors," Revision 1, October 2013.

2808 2809 19. NRC, RG 1.185, Standard Format and Content for Post-Shutdown Decommissioning Activities 2810 Report, Revision 1, June 2013.

2811 2812 20. NRC, RG 4.1, Radiological Environmental Monitoring for Nuclear Power Plants, Revision 2, 2813 June 2009.

2814 2815 21. NRC, RG 4.13, Environmental DosimetryPerformance Specifications, Testing, and Data 2816 Analysis, Revision 2, June 2019.

2817 2818 22. NRC, RG 4.15, Quality Assurance for Radiological Monitoring Programs (Inception Through 2819 Normal Operations to License Termination)Effluent Streams and the Environment, 2820 Revision 2, July 2007.

2821 2822 23. NRC, RG 4.20, Constraint on Releases of Airborne Radioactive Materials to the Environment 2823 for Licensees other than Power Reactors, Revision 1, April 2012.

2824 2825 24. NRC, RG 4.25, Assessment of Abnormal Radionuclide Discharges in Groundwater to the 2826 Unrestricted Area at Nuclear Power Plant Sites, Revision 0, March 2017.

2827 2828 25. NRC, Generic Letter 89-01, Implementation of Programmatic and Procedural Controls for 2829 Radiological Effluent Technical Specifications, January 31, 1989, ADAMS Accession 2830 No. ML031140051.

2831 2832 26. NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power 2833 Plants" issued October 1978, ADAMS Accession No. ML091050057.

2834 2835 27. NRC, NUREG-0016, Calculation of Releases of Radioactive Materials in Gaseous and Liquid 2836 Effluents from Boiling-Water Reactors: GALE-BWR 3.2 Code, Revision 1, January 1979, 2837 ADAMS Accession No. ML091910213, and Revision 2, July 2020, ADAMS Accession 2838 No. ML20213C728.

2839 2840 28. NRC, NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid 2841 Effluents from Pressurized Water Reactors: GALE-PWR 3.2 Code, Revision 1, April 1985, 2842 ADAMS Accession No. ML112720A411, and Revision 2, July 2020, ADAMS Accession 2843 No. ML20213C729.

2844 RG 1.21, Rev. 3, Page 63

2845 29. NRC, NUREG-0543, Methods for Demonstrating LWR Compliance with the EPA Uranium 2846 Fuel Cycle Standard (CFR Part 190), issued January 1980, ADAMS Accession 2847 No. ML081360410.

2848 2849 30. NRC, NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980, 2850 ADAMS Accession No. ML051400209.

2851 2852 31. NRC, NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard Radiological 2853 Effluent Controls for Pressurized Water Reactors, April 1991, ADAMS Accession 2854 No. ML091050061.

2855 2856 32. NRC, NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard Radiological 2857 Effluent Controls for Boiling Water Reactors, April 1991, ADAMS Accession 2858 No. ML091050059.

2859 2860 33. NRC, NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual 2861 (MARSSIM), Revision 1, August 2000.

2862 2863 34. NRC, NUREG-1576, Multi-Agency Radiological Laboratory Analytical Protocols Manual, 2864 July 2004, ADAMS Accession No. ML042310547, ML042310738, ML042320083.

2865 2866 35. NRC, NUREG-1757, Consolidated Decommissioning Guidance: Characterization, Survey, and 2867 Determination of Radiological Criteria, Volume 2, Revision 1, September 2006, ADAMS 2868 Accession No. ML063000243.

2869 2870 36. NRC, NUREG-1940, RASCAL 4: Description of Models and Methods, December 2012, 2871 ADAMS Accession No. ML13031A448.

2872 2873 37. NRC, NUREG-1940, RASCAL 4.3: Description of Models and Methods, Supplement 1, 2874 May 2015, ADAMS Accession No. ML15132A119.

2875 2876 38. NRC, NUREG/CR-6948, Integrated Ground-Water Monitoring Strategy for NRC-Licensed 2877 Facilities and Sites: Logic, Strategic Approach and Discussion, Volume 1, November 2007, 2878 ADAMS Accession No. ML073310297.

2879 2880 39. NRC, NUREG/CR-6805, A Comprehensive Strategy of Hydrogeology Modeling and 2881 Uncertainty Analysis for Nuclear Facilities and Sites, July 2003, ADAMS Accession 2882 No. ML032470827.

2883 2884 40. NRC, Regulatory Issue Summary (RIS) 2008-03, Return/Reuse of Previously Discharged 2885 Radioactive Effluents February 2008, ADAMS Accession No. ML072120368.

2886 2887 41. NRC International Policy Statement, ADAMS Accession No. ML14132A317.

2888 2889 42. NRC Management Directive and Handbook 6.6, Regulatory Guides ADAMS Accession No.

2890 ML16083A122.

2891 2892 43. IAEA, Radiation Protection of the Public and the Environment, GSG 8, Vienna, Austria, RG 1.21, Rev. 3, Page 64

2893 2018.13 2894 2895 44. IAEA, Dispersion of Radioactive Material in Air and Water and Consideration of Population 2896 Distribution in Site Evaluation for Nuclear Power Plants, Specific Safety Guide No. NS-G-3.2, 2897 Vienna, Austria, 2002.

2898 2899 45. IAEA, Regulatory Control of Radioactive Discharges to the Environment, GSG 9, Vienna, 2900 Austria, 2018.

2901 2902 46. IAEA, Environmental and Source Monitoring for Purposes of Radiation Protection, 2903 GSG N. RS-G-1.8, Vienna, Austria, 2005.

2904 2905 47. IAEA, Accident Monitoring Systems for Nuclear Power Plants, Nuclear Energy Series No. NP-2906 T-3.16, Vienna, Austria, 2015.

2907 2908 48. IAEA, Prevention and Mitigation of Groundwater Contamination from Radioactive Releases, 2909 TECDOC-482, Vienna, Austria, issued 1988.

2910 2911 49. IAEA, Remediation Process for Areas Affected by Past Activities and Accidents, IAEA Safety 2912 Guide No. WS-G-3.1, Vienna, Austria, 2007.

2913 2914 50. IAEA, Management of Waste Containing Tritium and Carbon-14, Technical Report Series 2915 Number 421, Vienna, Austria, 2004.

2916 2917 51. NRC, RG 1.21, Measuring, Evaluating, and Reporting Radioactive Material in Liquid and 2918 Gaseous Effluents and Solid Waste, Revision 1, June 1974.

2919 2920 52. NRC, RG 1.21, Measuring, Evaluating, and Reporting Radioactive Material in Liquid and 2921 Gaseous Effluents and Solid Waste, Revision 2, June 2009.

2922 2923 53. NRC, Contamination of Nonradioactive System and Resulting Potential for Unmonitored, 2924 Uncontrolled Release of Radioactivity to Environment, Inspection and Enforcement Bulletin 2925 No. 80-10, May 1980.

2926 2927 54. NRC, Results of the License Termination Rule Analysis, Commission Paper SECY-03-0069, 2928 May 23, 2003, ADAMS Accession No. ML030800158.

2929 2930 55. NRC, NUREG-1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73: Final Report, 2931 January 2013, ADAMS Accession No. ML13032A220.

2932 2933 56. NRC, NUREG/BR-0308, Effective Risk Communication, January 2004, ADAMS Accession 2934 No. ML040690412.

2935 2936 57. NRC, SRM-SECY-13-108, Staff RequirementsSECY-13-108Staff Recommendations for 2937 Addressing Remediation of Residual Radioactivity During Operations, December 20, 2013, 2938 ADAMS Accession No. ML13354B759.

13 Copies of IAEA documents may be obtained through their Web site: https://www.iaea.org/ or by writing the International Atomic Energy Agency, P.O. Box 100 Wagramer Strasse 5, A-1400 Vienna, Austria.

RG 1.21, Rev. 3, Page 65

2939 58. EPRI14 Report 1021104 Groundwater and Soil Remediation Guidelines for Nuclear Power 2940 Plants, (Proprietary report) issued December 2010.

2941 2942 59. EPRI Report 1023464, Groundwater and Soil Remediation Guidelines for Nuclear Power 2943 Plants (Public Edition) Final Report, July 2011.

2944 2945 60. NRC, NUREG/CR-6676, Probabilistic Dose Analysis Using Parameter Distributions Developed 2946 for RESRAD and RESRAD-BUILD Codes, July 2000, ADAMS Accession No. ML003741920.

2947 2948 61. NRC, NUREG/CR-6692, Probabilistic Modules for the RESRAD and RESRAD-BUILD 2949 Computer Codes, November 2000, ADAMS Accession No. ML003774030.

2950 2951 62. NRC, NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 2952 3.0 Computer Codes, December 2000, ADAMS Accession No. ML010090284.

2953 2954 63. NRC, NUREG/CR-7267, Default parameter Values and Distributions in RESRAD-ONSITE 2955 V7.2, RESRAD-BUILD V3.5 and RESRAD-OFFSITE 4.0, 2020, ADAMS Accession No.

2956 ML20279A652.

2957 2958 64. National Council on Radiation Protection and Measurements, Carbon-14 in the Environment, 2959 Report No. 81, Bethesda, MD, January 1985.

2960 2961 65. EPRI, Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, Technical 2962 Report 1021106, Palo Alto, CA, December 2010.

2963 2964 66. EPRI, "Carbon-14 Dose Calculation Methods at Nuclear Power Plants," Technical Report 2965 No. 1024827, Palo Alto, CA, April 2012.

2966 2967 67. ASTM D 3370 - 18, Standard Practices for Sampling Water from Flowing Process 2968 Streams ASTM D 3370 - 18, West Conshohocken, PA, 2007.15 2969 2970 68. American National Standards Institute (ANSI), Specification and Performance of On-Site 2971 Instrumentation for Continuously Monitoring Radioactivity in Effluents, ANSI N42.18-2004, 2972 New York, NY.

2973 2974 69. ANSI Instrumentation and Systems for Monitoring Radioactivity, ANSI N42.54-2018, New 2975 York, NY 2976 2977 70. ANSI /Health Physics Society, Sampling and Monitoring Releases of Airborne Radioactive 2978 Substances from the Stacks and Ducts of Nuclear Facilities, ANSI/HPS N13.1-2011, New York, 2979 NY.

2980 2981 71. NRC, Onsite Meteorological Programs, Safety Guide 23, February 17, 1972, ADAMS 2982 Accession No. ML020360030.

14 Copies of EPRI standards and reports may be obtained from EPRI, 3420 Hillview Ave., Palo Alto, CA 94304; telephone (800) 313-3774; https://www.epri.com 15 Copies of ASTM standards may be purchased from ASTM, 100 Barr Harbor Drive, P.O. Box C700, West Conshohocken, Pennsylvania 19428-2959; telephone (610) 832-9585. Purchase information is available through the ASTM Web site at http://www.astm.org.

RG 1.21, Rev. 3, Page 66

2983 2984 72. ANSI /ANS, Evaluation of Subsurface Radionuclide Transport at Commercial Nuclear Power 2985 Production Facilities, ANSI/ANS 2.17-2009, New York, NY.

2986 2987 73. EPRI Report No. 3002000546 Groundwater Protection Guidelines for Nuclear Power Plants:

2988 Revision 1, Electric Power Research Institute, Palo Alto, CA., October 2013.

2989 2990 74. ANSI, Radiation Protection Instrumentation Test and CalibrationAir Monitoring 2991 Instruments, ANSI N323C-2009, New York, NY.

2992 2993 75. D.G. Eisenhut, NRC, memorandum for Regional Administrators, Proposed Guidance for 2994 Calibration and Surveillance Requirements for Equipment Provided to Meet Item II.F.1, 2995 Attachments 1, 2, and 3, NUREG-0737, August 16, 1982, ADAMS Accession 2996 No. ML103420044.

2997 2998 76. NRC, NUREG/CR-5569, Proposed Guidance for Calibration and Surveillance Requirements to 2999 Meet Item II.F.1 of NUREG-0737, HPPOS-001 in Health Physics Positions Data Base, 3000 Revision 1, February 1994, ADAMS Accession No. ML093220108.

3001 3002 77. ANSI, Performance Specifications for Reactor Emergency Radiological Monitoring 3003 Instrumentation, ANSI N320-1978, New York, NY.

3004 3005 78. ICRP Publication 2, Report of Committee II on Permissible Dose for Internal Radiation, 3006 International Commission on Radiation Protection, Pergamon Press, Oxford, 1959.

3007 3008 79. Maiello, M., The Variations in Long Term TLD Measurements of Environmental Background 3009 Radiation at Locations in Southeastern New York State and Southern New Jersey, Health 3010 Physics, 72:915-922, June 1997.

3011 3012 80. ANSI /HPS, American National Standard for Dosimetry Personnel Dosimetry Performance 3013 Criteria for Testing, ANSI/HPS N13.11-2009, New York NY, January 13, 2009.

3014 3015 81. ANSI/HPS, Environmental DosimetryCriteria for System Design and Implementation, 3016 ANSI/HPS N13.37-2014, New York NY, April 8, 2014.

3017 3018 82. NRC, NUREG-1430, Standard Technical Specifications, Babcock and Wilcox Plants, April 3019 2012, ADAMS Accession No. ML12100A177 and ML12100A178.

3020 3021 83. NRC, NUREG-1431, Standard Technical Specifications, Westinghouse Plants, April 2012, 3022 ADAMS Accession No. ML12100A222 and ML12100AA288.

3023 3024 84. NRC, NUREG-1432, Standard Technical Specifications, Combustion Engineering Plants, 3025 April 2012, ADAMS Accession No. ML12102A165 and ML12102A169.

3026 3027 85. NRC, NUREG-1433, Standard Technical Specifications, General Electric BWR/4 Plants, 3028 April 2012, ADAMS Accession No. ML12024A192 and ML12104A193.

3029 3030 86. NRC, NUREG-1434. Standard Technical Specifications, General Electric BWR/6, April 2012, 3031 ADAMS Accession No. ML12104A195 and ML12104A196.

3032 RG 1.21, Rev. 3, Page 67

3033 87. ASTM, Standard Practice for Using Significant Digits in Test Data to Determine Conformance 3034 with Specifications, ASTM E-29, West Conshohocken, PA.

3035 3036 88. NRC, NUREG/CR-5569, Attention to Liquid Dilution Volumes in Semiannual Radioactive 3037 Effluent Release Reports, HPPOS-099, in Health Physics Positions Data Base, 3038 November 1984, ADAMS Accession No. ML093220108.

3039 3040 89. NEI, Industry Groundwater Protection InitiativeFinal Guidance Document, NEI 07-07, 3041 Revision 1, Washington, DC, February 26, 2019, ADAMS Accession No. ML20199M271.

3042 3043 90. NRC, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, 3044 Management Directive 8.4, September 2019, ADAMS Accession No. ML18093B087.

3045 3046 91. NRC, NUREG/CR-4007, Lower Limit of Detection: Definition and Elaboration of a Proposed 3047 Position for Radiological Effluent and Environmental Measurements, September 1984, ADAMS 3048 Accession No. ML16152A647.

3049 3050 92. NRC, NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey 3051 Instruments for Various Contaminants and Field Conditions, June 1998, ADAMS Accession 3052 No. ML20233A507.

3053 3054 93. ANSI, Traceability of Radioactive Sources to the National Institute of Standards and 3055 Technology (NIST) and Associated Instrument Quality Control, ANSI N42.22-2006, New York, 3056 NY.

3057 3058 94. NRC, NUREG/CR-5569, Definition of Unplanned Release, HPPOS-254, in Health Physics 3059 Positions Data Base, February 1994, ADAMS Accession No. ML093220108.

RG 1.21, Rev. 3, Page 68

3060 BIBLIOGRAPHY 3061 3062 U.S. Nuclear Regulatory Commission Documents 3063 3064 NUREG-Series Reports 3065 3066 U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports 3067 for Nuclear Power Plants, NUREG-0800, Section 2.3.5, Long-Term Atmosphere Dispersion Estimates 3068 for Routine Releases, Revision 3, Washington, DC, March 2007.

3069 3070 XOQDOQ: Program for the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power 3071 Stations, NUREG-0324, September 1977.

3072 3073 XOQDOQ: Computer Program for the Meteorological Evaluation of Routine Effluent Releases at 3074 Nuclear Power Stations, NUREG/CR2919, September 1982.

3075 3076 Regulatory Guides 3077 3078 U.S. Nuclear Regulatory Commission, Design Guidance for Radioactive Waste Management Systems, 3079 Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, Regulatory 3080 Guide 1.143, Revision 2, November 2001.

3081 3082 U.S. Environmental Protection Agency Documents 3083 3084 U.S. Code of Federal Regulations, National Primary Drinking Water Regulations, Part 141, Chapter 1, 3085 Title 40, Protection of Environment.

3086 3087 National Standards and Industry Reports 3088 3089 ANSI, Performance Criteria for Radiobioassay, ANSI N13.30-1996, New York, NY.

3090 3091 ANSI/ANS, Determining Meteorological Information at Nuclear Facilities, ANSI/ANS 3.11-2005, 3092 New York, NY, January 2005.

3093 3094 ANSI, Calibration and Use of Germanium Spectrometers for the Measurement of Gamma-Ray Emission 3095 Rates of Radionuclides, ANSI N42.14-1999, New York, NY.

3096 3097 ANSI/National Conference of State Legislatures (NCSL), American National Standard for Expressing 3098 UncertaintyU.S. Guide to the Expression of Uncertainty in Measurement, ANSI/NCSL Z540-2-1997 3099 (reapproved 2002), New York, NY.

3100 3101 NIST, Guidelines for Evaluating and Expressing the Uncertainty of NIST Measurement Results, 3102 Technical Note 1297, Gaithersburg, MD, September 1994.

3103 3104 EPRI, Groundwater Monitoring Guidance for Nuclear Power Plants, Report No. 1011730, Palo Alto, 3105 CA, September 2005.

3106 3107 EPRI, Groundwater Protection Guidelines for Nuclear Power Plants, Rev. 1 Report No. 3002000546, 3108 Palo Alto, CA, November 2007.

RG-1.21, Rev. 3, Page 69

APPENDIX ATABLES Table A Gaseous EffluentsSummation of All Releases SUMMATION OF UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL UNCERTAINTY ALL RELEASES Fission and Ci Activation Gases Iodines (Halogens) Ci Particulates Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-1

Table A-1A - Gaseous EffluentsGround-Level ReleaseBatch Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Ar-41 Ci Kr-85 Ci Kr-85m Ci Kr-87 Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci Xe-138 Ci (List Others) Ci Total Ci Iodines/

UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Halogens I-131 Ci I-132 Ci I-133 Ci I-134 Ci I-135 Ci Total Ci Particulates UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Cs-134 Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-2

Table A-1B - Gaseous EffluentsGround-Level ReleaseContinuous Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Ar-41 Ci Kr-85 Ci Kr-85m Ci Kr-87 Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci Xe-138 Ci (List Others)

Total Ci Iodines/

UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Halogens I-131 Ci I-132 Ci I-133 Ci I-134 Ci I-135 Ci Total Ci Particulates UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Cs-134 Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-3

Table A-1C - Gaseous EffluentsElevated ReleaseBatch Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Ar-41 Ci Kr-85 Ci Kr-85m Ci Kr-87 Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci Xe-138 Ci (List Others) Ci Total Ci Iodines/

UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Halogens I-131 Ci I-132 Ci I-133 Ci I-134 Ci I-135 Ci Total Ci Particulates UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Cs-134 Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-4

Table A-1D - Gaseous EffluentsElevated ReleaseContinuous Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Ar-41 Ci Kr-85 Ci Kr-85m Ci Kr-87 Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci Xe-138 Ci (List Others) Ci Total Ci Iodines/

UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Halogens I-131 Ci I-132 Ci I-133 Ci I-134 Ci I-135 Ci Total Ci Particulates UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Cs-134 Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-5

Table A-1E - Gaseous EffluentsMixed Mode ReleaseBatch Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Ar-41 Ci Kr-85 Ci Kr-85m Ci Kr-87 Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci Xe-138 Ci (List Others) Ci Total Ci Iodines/

UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Halogens I-131 Ci I-132 Ci I-133 Ci I-134 Ci I-135 Ci Total Ci Particulates UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Cs-134 Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-6

Table A-1F - Gaseous EffluentsMixed Mode ReleaseContinuous Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Ar-41 Ci Kr-85 Ci Kr-85m Ci Kr-87 Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci Xe-138 Ci (List Others) Ci Total Ci Iodines/

UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Halogens I-131 Ci I-132 Ci I-133 Ci I-134 Ci I-135 Ci Total Ci Particulates UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Cs-134 Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci C-14 Ci RG 1.21, Rev. 3, Appendix A, Page A-7

Table A Liquid EffluentsSummation of All Releases SUMMATION OF ALL LIQUID UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL UNCERTAINTY (%)

RELEASES Fission and Activation Ci Products (excluding tritium, noble gases, C-14 and gross alpha)

Tritium Ci Dissolved and Entrained Ci Gases Gross Alpha Ci Volume of Primary Liters System Liquid Effluent (before dilution)

Dilution Water Used for Liters Above Volume of Secondary or Liters Balance-of-Plant Liquid Effluent (e.g., low-activity or unprocessed)

(before dilution)

Quarterly Dilution Liters Water Used for Above Average Stream Flow m3/s RG 1.21, Rev. 3, Appendix A, Page A-8

Table A-2A - Liquid EffluentsBatch Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Products Cr-51 Ci Mn-54 Ci Fe-55 Ci Fe-59 Ci Co-57 Ci Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Nb-95 Ci Ag-110m Ci Sn-113 Ci Sb-124 Ci Sb-125 Ci I-131 Ci I-133 Ci I-135 Ci Cs-134 Ci Cs-137 Ci (List Others) Ci Total Ci RG 1.21, Rev. 3, Appendix A, Page A-9

Table A-2A - Liquid EffluentsBatch Mode (continued)

Dissolved and UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Entrained Gases Kr-85 Ci Kr-85m Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci RG 1.21, Rev. 3, Appendix A, Page A-10

Table A-2B - Liquid EffluentsContinuous Mode Fission and Activation UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Products Cr-51 Ci Mn-54 Ci Fe-55 Ci Fe-59 Ci Co-57 Ci Co-58 Ci Co-60 Ci Sr-89 Ci Sr-90 Ci Nb-95 Ci Ag-110m Ci Sn-113 Ci Sb-124 Ci Sb-125 Ci I-131 Ci I-133 Ci I-135 Ci Cs-134 Ci Cs-137 Ci (List Others) Ci Total Ci RG 1.21, Rev. 3, Appendix A, Page A-11

Table A-2B - Liquid EffluentsContinuous Mode (continued)

Dissolved and Entrained UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL Gases Kr-85 Ci Kr-85m Ci Kr-88 Ci Xe-131m Ci Xe-133 Ci Xe-133m Ci Xe-135 Ci Xe-135m Ci (List Others) Ci Total Ci Tritium Ci Gross Alpha Ci RG 1.21, Rev. 3, Appendix A, Page A-12

Table A Solid Waste and Irradiated Fuel Shipments A. SOLID RADIOACTIVE WASTE SHIPPED FROM THE UNIT (not irradiated fuel)

ACTIVITY OF NUMBER OF VOLUME TYPE OF WASTE MAJOR NUCLIDES SHIPMENTS (m3)

(Ci)

Wet radioactive waste (e.g., spent resins, filters, sludges, etc.)

Dry radioactive waste (e.g., trash, paper, discarded protective clothing, etc.)

Activated or contaminated radioactive material (e.g., equipment or bulk radioactive material)

Other radioactive waste (waste not included in the above categories and waste not excepted per Section 6 of this RG.)

B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation Destination RG 1.21, Rev. 3, Appendix A, Page A-13

Table A Dose Limits16, per Technical Specifications (based on fractions of 10 CFR Part 50, Appendix I)

QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 YEARLY Liquid Effluent 1.5 mrem 1.5 mrem 1.5 mrem 1.5 mrem 3 mrem Dose Limit, Total Body Total Body Dose

% of Dose Limit Liquid Effluent 5 mrem 5 mrem 5 mrem 5 mrem 10 mrem Dose Limit, Any Organ Organ Dose

% of Dose Limit Gaseous Effluent 5 mrad 5 mrad 5 mrad 5 mrad 10 mrad Dose Limit, Gamma Air Gamma Air Dose

% of Dose Limit Gaseous Effluent 10 mrad 10 mrad 10 mrad 10 mrad 20 mrad Dose Limit, Beta Air Beta Air Dose

% of Dose Limit Gaseous Effluent 7.5 mrem 7.5 mrem 7.5 mrem 7.5 mrem 15 mrem Organ Dose Limit (iodine, tritium, particulates with >8-day half-life)

Gaseous Effluent Organ Dose (iodine, tritium, particulates with > 8-day half-life)

% of Dose Limit 16 Doses based on quarterly and annual limits RG 1.21, Rev. 3, Appendix A, Page A-14

1 Table A EPA 40 CFR Part 190 Dose Limits to an Individual in the Unrestricted Area 2

WHOLE BODY THYROID ANY OTHER ORGAN Dose Limit 25 mrem 75 mrem 25 mrem Dose17

% of Dose Limit 3

4 17 Dose from current year effluent discharges.

RG 1.21, Rev. 3, Appendix A, Page A-15

5 Table A-6. Supplemental Information 6

7 1. Abnormal Releases and Abnormal Discharges (e.g., leaks and spills) 8 9 2. Nonroutine, Planned Discharges (e.g., pumping of leaks and spills for remediation, results of 10 groundwater monitoring to quantify effluent releases to the offsite environment) 11 12 3. Radioactive Waste Treatment System Changes 13 14 4. Annual Land Use Census Changes 15 16 5. Effluent Monitor Instrument Inoperability 17 18 6. ODCM Changes 19 20 7. Process Control Program Changes 21 22 8. Errata/Corrections to Previous ARERRs 23 24 9. Other (narrative description of other information that is provided to the NRC, such as in the 25 ARERR or ISFSI reports).

RG 1.21, Rev. 3, Appendix A, Page A-16