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)

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U.S. NUCLEAR REGULATORY COMMISSION REGULATORY GUIDE RG 1.21 Issue Date: Month 2021 Technical Lead: Steven Garry 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.

MEASURING, EVALUATING, AND REPORTING 1

RADIOACTIVE MATERIAL IN LIQUID AND GASEOUS 2

EFFLUENTS AND SOLID WASTE 3

4 A. INTRODUCTION 5

6 Purpose 7

8 This regulatory guide (RG) describes methods the staff of the U.S. Nuclear Regulatory 9

Commission (NRC) considers acceptable for the following uses:

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

19 20 This guide incorporates the risk-informed principles of the Reactor Oversight Process. A 21 risk-informed, performance-based approach to regulatory decision making combines the risk-informed 22 and performance-based elements discussed in the staff requirements memorandum to SECY-98-144, 23 Staff RequirementsSECY-98-144White Paper on Risk-Informed and Performance-Based 24 Regulation, dated February 24, 1999 (Ref. 3).

25 26 Applicability 27 28 This RG is a Division 1, Power Reactors RG, which applies to nuclear power plant licensees 29 and applicants subject to Title 10 of the Code of Federal Regulations (10 CFR) Part 20, Standards for 30 Protection Against Radiation, This RG is also applicable to specific and general licensees under 10 Part 31 72 for storage of spent fuel.

32 This includes licenses issued under the following regulations:

33 34

RG 1.21, Rev. 3, Page 2 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities (Ref. 4), applies to 35 the licensing of production and utilization facilities.

36 37 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 5),

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

40 41 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 42 High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste (Ref. 6),

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

45 46 Applicable Regulations 47 48 The following regulations establish the regulatory basis for the radiological effluent control 49 program:

50 51 10 CFR Part 20, Standards for Protection Against Radiation 52 53 o 10 CFR 20.1003, Definitions, defines terminology that is used in the regulations and in 54 this regulatory guide.

55 56 o 10 CFR 20.1301, Dose limits for individual members of the public, establishes 57 radiation dose limits for individual members of the public.

58 59 o 10 CFR 20.1302, Compliance with dose limits for individual members of the public, 60 requires licensees to perform surveys of radiation levels in unrestricted and controlled 61 areas and radioactive materials in effluents released to unrestricted and controlled areas to 62 demonstrate compliance with the dose limits for individual members of the public.

63 64 o 10 CFR 20.1402, Radiological criteria for unrestricted use, establishes acceptance 65 criteria for license termination to achieve the sites unrestricted use status after 66 decommissioning.

67 68 o 10 CFR 20.1501, General, establishes requirements for performing radiological 69 surveys.

70 71 o 10 CFR 20.2001, General requirements (for waste disposal), establishes methods for 72 disposing of licensed material.

73 74 o 10 CFR 20.2103, Records of surveys, requires licensees to maintain records of surveys 75 and calibrations.

76 77 o 10 CFR 20.2107, Records of dose to individual members of the public, requires 78 licensees to maintain records that demonstrate compliance with dose limits for members 79 of the public.

80 81 o 10 CFR 20.2108, Records of waste disposal, requires licensees to maintain records of 82 the disposal of licensed material.

83 84

RG 1.21, Rev. 3, Page 3 o 10 CFR Part 20, Appendix B, Annual Limits on Intakes (ALIs) and Derived Air 85 Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent 86 Concentrations; Concentrations for Release to Sewage, establishes intake limits and 87 airborne and liquid concentration limits for occupational exposure and member of the 88 public exposure.

89 90 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities 91 92 o 10 CFR 50.34a, Design objectives for equipment to control releases of radioactive 93 material in effluentsnuclear power reactors, establishes numerical guides for design 94 objectives and limiting conditions of operation to control radioactive effluents.

95 96 o 10 CFR 50.36a, Technical specifications on effluents from nuclear power reactors, 97 requires licensees to establish technical specifications with operating procedures and 98 controls be established and followed and that the radioactive waste system be maintained 99 and used.

100 101 o 10 CFR 50.75, Reporting and record keeping for decommissioning planning, 102 paragraph (g,) requires licensees to keep records of information important to 103 decommissioning.

104 105 o 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, 106 General Design Criterion (GDC) 60, Control of Releases of Radioactive Materials to the 107 Environment, specifies that the nuclear power unit design shall include means to control 108 suitably liquid and gaseous effluents and solid waste.

109 110 o 10 CFR Part 50, Appendix A, GDC 64, Monitoring Radioactivity Releases, specifies 111 that means shall be provided for monitoring the reactor containment atmosphere, spaces 112 containing components for recirculation of loss-of-coolant fluids, effluent discharge paths 113 and the plant environs for radioactivity that may be released from normal operations, 114 anticipated operational occurrences, and from postulated accidents.

115 116 o 10 CFR Part 50, Appendix I, Numerical Guides for Design Objectives and Limiting 117 Conditions for Operation to Meet the Criterion As Low As Is Reasonably Achievable 118 (ALARA) for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor 119 Effluents, establishes design objectives for meeting the requirements of 10 CFR 50.34a.

120 121 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants 122 123 o 10 CFR 52.0, Scope, requires Part 52 licensees to comply with all requirements in 124 10 CFR Chapter I that are applicable, which includes, for example, 10 CFR Part 20 as 125 discussed above.

126 127 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, 128 High Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste 129 130 o 10 CFR 72.44(d) requires that each specific license must include technical specifications 131 that establishes limits on the release of radioactive materials and the ALARA objectives 132 for effluents and that require establishment of an environmental monitoring program to 133 ensure compliance with those limits 134

RG 1.21, Rev. 3, Page 4 135 o 10 CFR 72.104, Criteria for radioactive materials in effluents and direct radiation from 136 an ISFSI or MRS, establishes dose limits to any real individual (excluding occupational 137 exposures) beyond the Part 72 controlled area (as defined in 10 CFR 72.3 and meeting 138 the minimum size requirements in 72.106(b))

139 140 o 10 CFR 72.126, Criteria for radiological protection, requires radiation protection 141 systems be provided with effluent and direct radiation monitoring systems and controls to 142 limit releases to ALARA under normal conditions and control releases under accident 143 conditions and ensure limits relating to releases to the general environment will not be 144 exceeded 145 146 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operations 147 148 o 40 CFR 190.10, Standards for normal operation, establishes standards for normal 149 operations and annual dose equivalent standards and limits on the total quantity of 150 radioactive materials entering the environment from the entire uranium fuel cycle.

151 152 o 40 CFR 190.11, Variances for unusual operations, establishes variances (allowances) 153 for unusual operations where the standards in 40 CFR 190.10 may be exceeded.

154 155 40 CFR Part 191, Environmental Radiation Protection Standards for Management and Disposal 156 of Spent Nuclear Fuel and Transuranic Radioactive Wastes (Ref. 7) 157 158 o 40 CFR 191.03(a), Standards, establishes standards for the management and storage of 159 spent nuclear fuel or transuranic radioactive wastes.

160 161 162 Related Guidance 163 164 RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants (Ref. 8), provides 165 guidance for an onsite meteorological measurements program.

166 167 RG 1.97, Revisions 0, 1, 2 and 3, Instrumentation for Light-Water-Cooled Nuclear Power Plants 168 to Assess Plant and Environs Conditions During and Following an Accident, issued 169 December 1975, August 1977, and December 1980, and May 1983, respectively (Ref. 9),

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

172 173 RG 1.97, Revision 4, Criteria for Accident Monitoring Instrumentation for Nuclear Power 174 Plants, issued June 2006 (Ref. 10), endorses (with certain clarifying regulatory positions) the 175 Institute of Electrical and Electronics Engineers (IEEE) Standard (Std.) 497-2002, IEEE 176 Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating 177 Stations (Ref. 11).

178 179 RG 1.97, Revision 5, Criteria for Accident Monitoring Instrumentation for Nuclear Power 180 Plants, issued April 2019 (Ref. 12), endorses, with exceptions and clarifications, IEEE 181 Std. 497-2016, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear 182 Power Generating Stations (Ref. 13).

183 184

RG 1.21, Rev. 3, Page 5 RG 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for 185 the Purpose of Demonstrating Compliance with 10 CFR Part 50, Appendix I (Ref. 14), describes 186 basic features of calculational models and assumptions used for the estimation of doses to the 187 public.

188 189 RG 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents 190 in Routine Releases from Light-Water-Cooled Reactors (Ref. 15), describes models and 191 assumptions for the estimation of atmospheric dispersion of gaseous effluent releases.

192 193 RG 1.112, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents 194 from Light-Water-Cooled Power Reactors, (Ref. 16), provides acceptable methods for applicants 195 to construct a nuclear power reactor to calculate realistic radioactive source terms for use in 196 evaluating radioactive waste treatment systems to determine whether the design objectives of 197 10 CFR Part 50, Appendix I, are met, and to assess the environmental impact of radioactive 198 effluents.

199 200 RG 1.113, Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor 201 Releases for the Purpose of Implementing Appendix I (Ref. 17), describes general approaches 202 for the analysis of releases of liquid effluents into surface water bodies.

203 204 RG 1.184, "Decommissioning of Nuclear Power Reactors" (Ref. 18), provides guidance that 205 during decommissioning, Technical Specifications require operational procedures for the control 206 of effluent releases and submittal of annual effluent reports as specified by 10 CFR 50.36a.

207 208 RG 1.185, Standard Format and Content for Post-Shutdown Decommissioning Activities 209 Report (Ref. 19), identifies information licensees should provide to NRC and the public of the 210 licensees expected decommissioning activities and schedule.

211 212 RG 4.1, Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants 213 (Ref. 20), describes acceptable programs for establishing and conducting an environmental 214 monitoring program.

215 216 RG 4.13, Environmental DosimetryPerformance Specifications, Testing, and Data Analysis 217 (Ref. 21), provides specifications for environmental dosimetry and methods of analyzing 218 dosimetry to determine dose to members of the public.

219 220 RG 4.15, Quality Assurance for Radiological Monitoring Programs (Inception Through Normal 221 Operations to License Termination)Effluent Streams and the Environment (Ref. 22), describes 222 design and implementation programs to ensure the quality of the results of measurements of 223 radioactive materials in the effluents from, and environment outside of, facilities that process, 224 use, or store radioactive materials.

225 226 RG 4.20, Constraint on Releases of Airborne Radioactive Materials to the Environment for 227 Licensees other than Power Reactors (Ref. 23), provides guidance for meeting the constraint on 228 airborne emissions of radioactive material as described in 10 CFR 20.1101(d).

229 230 RG 4.25, Assessment of Abnormal Radionuclide Discharges in Groundwater to the Unrestricted 231 Area at Nuclear Power Plant Sites (Ref. 24), describes an approach that is acceptable for use in 232 assessing abnormal discharges of radionuclides in groundwater from the subsurface to the 233 unrestricted area at nuclear power plant sites.

234

RG 1.21, Rev. 3, Page 6 235 Generic Letter (GL) 89-01, Guidance for the Implementation of Programmatic Controls for 236 Radiological Effluent Technical Specifications in the Administrative Controls Section of 237 Technical Specifications and the Relocation of Procedural Details to the Offsite Dose Calculation 238 Manual or Process Control Program, dated January 31, 1989 (Ref. 25), provides guidance for the 239 preparation of a license amendment request to relocate programmatic controls for radioactive 240 effluents and for radiological environmental monitoring from technical specifications to the 241 licensee-controlled Offsite Dose Calculation Manual (ODCM) or equivalent document.

242 243 NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power 244 Plants" issued October 1978 (Ref. 26), is one of the bases documents for the Radioactive Effluent 245 Controls Program in Standard Technical Specifications (section 5.5.4).

246 247 NUREG-0016, Revision 1 and Revision 2, Calculation of Releases of Radioactive Materials in 248 Gaseous and Liquid Effluents from Boiling-Water Reactors (GALE-BWR 3.2 Code), issued 249 January 1979 and July 2020, respectively (Ref. 27), is a computerized mathematical model for 250 calculating the release of radioactive materials in gaseous and liquid effluents from boiling-water 251 reactors (BWRs).

252 253 NUREG-0017, Revision 1 and Revision 2, Calculation of Releases of Radioactive Materials in 254 Gaseous and Liquid Effluents from Pressurized-Water Reactors (GALE-PWR 3.2 Code), issued 255 April 1985 and July 2020, respectively (Ref. 28), is a computerized mathematical model for 256 calculating the release of radioactive materials in gaseous and liquid effluents from 257 pressurized-water reactors (PWRs).

258 259 NUREG-0543, Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel 260 Cycle Standard (CFR Part 190), 1980 (Ref. 29) explains the rationale for using Appendix I to 261 demonstrate compliance with 40 CFR 190 and methods for demonstrating compliance when 262 radioactive effluents exceed Appendix I numerical guidance.

263 264 NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980 265 (Ref. 30), provides specific items that were approved by the NRC Commission following the 266 accident at Three Mile Island Nuclear Station (TMI) for implementation at reactors.

267 268 NUREG-1301, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 269 Controls for Pressurized Water Reactors, issued April 1991 (Ref. 31), provides the PWR effluent 270 controls that may be removed from technical specifications and incorporated into the licensees 271 ODCM (or equivalent).

272 273 NUREG-1302, Offsite Dose Calculation Manual Guidance: Standard Radiological Effluent 274 Controls for Boiling Water Reactors, issued April 1991 (Ref. 32), provides the BWR effluent 275 controls that may be removed from technical specifications and incorporated into the licensees 276 ODCM (or equivalent).

277 278 NUREG-1575, Rev. 1, Multi-Agency Radiation Survey and Site Investigation Manual 279 (MARSSIM) (Ref. 33), provides information on planning, conducting, evaluating, and 280 documenting building surface and surface soil final status radiological surveys for demonstrating 281 compliance with dose or risk-based regulations or standards.

282 283

RG 1.21, Rev. 3, Page 7 NUREG-1576, Multi-Agency Radiological Laboratory Analytical Protocols Manual (Ref. 34),

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

286 287 NUREG-1757, Volume 2, Revision 1, Consolidated Decommissioning Guidance:

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

291 292 NUREG-1940, RASCAL 4: Description of Models and Methods, issued December 2012 293 (Ref. 36), provides a description of an emergency response consequence assessment tool 294 including models and methods for source term calculations, atmospheric dispersion and 295 deposition, and dose calculations.

296 297 NUREG-1940, Supplement 1, RASCAL 4.3: Description of Models and Methods, issued 298 May 2015 (Ref. 37), describes the Radiological Assessment System for Consequence Analysis 299 (RASCAL) models and methods for source term calculations, atmospheric dispersion and 300 deposition, and dose calculations for accident analysis.

301 302 NUREG/CR-6948, Integrated Ground-Water Monitoring Strategy for NRC-Licensed Facilities 303 and Sites issued November 2007 (Ref. 38), presents a framework for assessing what, where, 304 when, and how to monitor contamination in groundwater.

305 306 NUREG/CR-6805, A Comprehensive Strategy of Hydrogeology Modeling and Uncertainty 307 Analysis for Nuclear Facilities and Sites, issued July 2003 (Ref. 39), describes a strategy for a 308 systematic and comprehensive approach to hydrogeologic conceptualization, model development, 309 and predictive uncertainty analysis.

310 311 Purpose of Regulatory Guides 312 313 The NRC issues RGs to describe methods that are acceptable to the staff for implementing 314 specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific 315 issues or postulated events, and to describe information that the staff needs in its review of applications 316 for permits and licenses. RGs are not NRC regulations and compliance with them is not required.

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

319 320 Paperwork Reduction Act 321 322 This RG provides voluntary guidance for implementing the mandatory information collections in 323 10 CFR Parts 20, 50, 52, 72, that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et.

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

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

331 oira_submission@omb.eop.gov.

332 333

RG 1.21, Rev. 3, Page 8 Public Protection Notification 334 335 The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of 336 information unless the document requesting or requiring the collection displays a currently valid OMB 337 control number.

338 339

RG 1.21, Rev. 3, Page 9 TABLE OF CONTENTS 340 341 A. INTRODUCTION.................................................................................................................................. 1 342 B. DISCUSSION....................................................................................................................................... 11 343 Reason for Revision................................................................................................................................ 11 344 Background............................................................................................................................................. 11 345 Objectives of the Radiological Effluent Controls Program.................................................................... 12 346 C. STAFF REGULATORY GUIDANCE................................................................................................. 15 347

1.

Effluent Monitoring.................................................................................................................... 15 348 1.1 Effluent Monitoring Programs................................................................................................ 15 349 1.2 Release Points for Effluent Monitoring.................................................................................. 15 350 1.3 Monitoring a Significant Release Point.................................................................................. 16 351 1.4 Monitoring a Less-Significant Release Point.......................................................................... 16 352 1.5 Monitoring Leaks and Spills................................................................................................... 17 353 1.6 Monitoring Continuous Releases............................................................................................ 19 354 1.7 Monitoring Batch Releases..................................................................................................... 20 355 1.8 Principal Radionuclides for Effluent Monitoring................................................................... 20 356 1.9 Carbon-14............................................................................................................................... 22 357 1.10 Return/Reuse of Previously Discharged Radioactive Effluents.............................................. 23 358 1.11 Abnormal Releases and Abnormal Discharges....................................................................... 23 359

2.

Effluent Sampling....................................................................................................................... 24 360 2.1 Representative Sampling......................................................................................................... 24 361 2.2 Sampling Liquid Radioactive Waste....................................................................................... 25 362 2.3 Sampling Gaseous Radioactive Waste.................................................................................... 25 363 2.4 Sampling Bias......................................................................................................................... 25 364 2.5 Composite Sampling............................................................................................................... 26 365 2.6 Sample Preparation and Preservation...................................................................................... 26 366 2.7 Short-Lived Radionuclides and Decay Corrections................................................................ 26 367

3.

Effluent Dispersion (Meteorology and Hydrology).................................................................... 26 368 3.1 Meteorological Data................................................................................................................ 26 369 3.2 Atmospheric Dispersion (Transport and Diffusion)............................................................... 27 370 3.3 Release Height........................................................................................................................ 27 371 3.4 Aquatic Dispersion (Surface Waters)...................................................................................... 28 372 3.5 Spills and Leaks to the Ground Surface.................................................................................. 28 373 3.6 Spills and Leaks to Groundwater............................................................................................ 28 374

4.

Quality Assurance....................................................................................................................... 30 375 4.1 Quality Assurance Programs................................................................................................... 30 376 4.2 Quality Control Checks........................................................................................................... 31 377 4.3 Surveillance Frequencies........................................................................................................ 31 378 4.4 Procedures............................................................................................................................... 31 379 4.5 Calibration of Laboratory Equipment and Routine Effluent Radiation Monitors................... 31 380 4.6 Calibration of Measuring and Test Equipment....................................................................... 32 381 4.7 Calibration Frequency............................................................................................................. 32 382 4.8 Measurement Uncertainty....................................................................................................... 32 383 4.9 Calibration of Accident-Range Radiation Monitors and Accident-Range Effluent Monitors 32 384

5.

Dose Assessments for Individual Members of the Public.......................................................... 34 385 5.1 Bounding Assessments........................................................................................................... 35 386 5.2 Individual Members of the Public........................................................................................... 36 387 5.3 Occupancy Factors.................................................................................................................. 36 388 5.4 10 CFR Part 50, Appendix I.................................................................................................... 36 389

RG 1.21, Rev. 3, Page 10 5.5 10 CFR 20.1301(a) through (c)............................................................................................... 37 390 5.6 10 CFR 20.1301(e).................................................................................................................. 37 391 5.7 Dose Assessments for 10 CFR Part 50, Appendix I............................................................... 38 392 5.8 Dose Assessments for 10 CFR 20.1301(e)............................................................................. 39 393 5.9 Dose Calculations................................................................................................................... 40 394

6.

Solid Radioactive Waste Released from the Unit...................................................................... 40 395

7.

Reporting Errata in Effluent Release Reports............................................................................. 41 396 7.1 Examples of Small Errors....................................................................................................... 41 397 7.2 Reporting Small Errors........................................................................................................... 41 398 7.3 Examples of Large Errors....................................................................................................... 41 399 7.4 Reporting Large Errors........................................................................................................... 42 400

8.

Changes to Effluent and Environmental Programs..................................................................... 43 401

9.

Format and Content of the Annual Radioactive Effluent Release Report...................................... 43 402 9.1 Gaseous Effluents................................................................................................................... 44 403 9.2 Liquid Effluents...................................................................................................................... 46 404 9.3 Solid Waste Shipments Released from the Unit (per Standard Technical Specifications)..... 47 405 9.4 Dose Assessments................................................................................................................... 48 406 9.5 Supplemental Information....................................................................................................... 48 407 D. IMPLEMENTATION........................................................................................................................... 52 408 GLOSSARY............................................................................................................................................... 53 409 REFERENCES........................................................................................................................................... 62 410 BIBLIOGRAPHY....................................................................................................................................... 69 411 APPENDIX ATABLES............................................................................................................................ 1 412 413 414

RG 1.21, Rev. 3, Page 11 B. DISCUSSION 415 416 Reason for Revision 417 418 This revision of RG 1.21 (Revision 3):

419 Provides guidance and acceptable methods for calibration of accident range radiation monitors 420 and accident range effluent monitors, 421 Revises guidance on recommendations for reviewing and updating long-term, annual average /Q 422 and D/Q values, 423 Clarifies reporting requirements for low level radioactive waste (LLW) shipments, specifically 424 that the report includes the waste shipped from the unit (plant site), and that waste classification 425 does not need to be reported when shipped from the unit (plant site) to a waste processor, 426 Clarifies the existing guidance in NUREG 1301 and NUREG 1302 that environmental monitoring 427 for iodine (I) -131 in drinking water should be performed if a prospective dose evaluation of the 428 annual thyroid dose from I-131 to a person in any age group from the drinking water route of 429 exposure is greater than one mrem.

430 Clarifies the existing process as currently described in Technical Specifications for making 431 changes to effluent and environmental programs, and, 432 Incorporates the existing Regulatory Issue Summary 2008-03, Return/Reuse of Previously 433 Discharged Radioactive Effluents (Ref. 40).

434 435

Background

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

442 443 (1)

RG 4.1 444 (2)

RG 4.15 445 (3)

RG 1.109 446 (4)

NUREG-1301 447 (5)

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

455 456 RG 1.21 addresses the measuring, evaluating, and reporting of effluent releases, solid radioactive 457 waste shipments, and public dose from nuclear power plants. The guide describes the important concepts 458 in planning and implementing an effluent and solid radioactive waste program. Concepts covered include 459 meteorology, release points, monitoring methods, identification of principal radionuclides, unrestricted 460 area boundaries, continuous and batch release methods, representative sampling, composite sampling, 461 radioactivity measurements, decay corrections, quality assurance (QA), solid radioactive waste shipments, 462

RG 1.21, Rev. 3, Page 12 and public dose assessments. The dose to occupational workers, including contributions from activities 463 associated with effluent programs (such as LLW processing, storage, and shipping, as well as dose from 464 handling resins and filters for gaseous and liquid radioactive waste), is occupational dose associated with 465 the licensed operation and is not included in RG 1.21.

466 467 RG 4.1 addresses the environmental monitoring program. The guide discusses principles and 468 concepts important to environmental monitoring at nuclear power plants. The RG provides guidance on 469 both the preoperational and operational Radiological Environmental Monitoring Programs for the 470 routinely monitored exposure pathways (inhalation, ingestion, and direct radiation). The guide defines 471 the sampling media and sampling frequency, and the methods of comparing environmental measurements 472 to effluent releases in the Annual Radiological Environmental Operating Report (AREOR).

473 474 RG 4.15 provides the basic principles of QA in all types of radiological monitoring programs for 475 effluent streams and the environment. The guide provides principles for structuring organizational lines 476 of communication and responsibility, using qualified personnel, implementing standard operating 477 procedures, defining data quality objectives (DQOs), performing quality control (QC) checking for 478 sampling and analysis, auditing the process, and taking corrective actions.

479 480 RG 1.109 provides the detailed implementation guidance for demonstrating that radioactive 481 effluents conform to ALARA design objectives of 10 CFR Part 50, Appendix I. The RG describes 482 calculational models and parameters for estimating dose from effluent releases, including the dispersion 483 of the effluent in the atmosphere and surface water bodies.

484 485 NUREG-1301 and NUREG-1302 provide the detailed implementation guidance by describing 486 effluent and environmental monitoring programs. These NUREGs provide guidance on meeting effluent 487 monitoring and environmental sampling requirements, surveillance requirements for effluent monitors, 488 types of monitors and samplers, sampling and analysis frequencies, types of analysis and radionuclides 489 analyzed, lower limits of detection (LLDs), specific environmental media to be sampled, and reporting 490 and program evaluation and revision.

491 492 Objectives of the Radiological Effluent Controls Program 493 494 The requirements for the radiological effluent control program are in 10 CFR Part 20 and the 495 technical specifications that are part of a license, including limitations on dose conforming to 496 10 CFR Part 50, Appendix I. In addition, a facilitys technical specifications describe specific regulatory 497 requirements. Licensees can use these regulatory requirements and the RG 1.21 regulatory guidance as a 498 basis for establishing the radiological effluent control program. The radiological effluent control program 499 for a nuclear power plant has the following six basic objectives, which are also reflected in 500 10 CFR 50.36a and in site-specific Technical Specifications:

501 502 (1)

Ensure that effluent instrumentation has the functional capability to measure and analyze effluent 503 discharges.

504 505 (2)

Ensure that effluent treatment systems are used to reduce effluent discharges to ALARA levels.

506 507 (3)

Establish instantaneous release-rate limitations on the concentrations of radioactive material.

508 509 (4)

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

511 512 (5)

Measure, evaluate, and report the quantities of radioactivity in gaseous effluents, liquid effluents, 513

RG 1.21, Rev. 3, Page 13 and solid radioactive waste shipments.

514 515 (6)

Evaluate the dose to members of the public.

516 517 As required by technical specifications, Part 50 and Part 52 licensees must submit the Annual 518 Radioactive Effluent Release Report (ARERR) before May 1 and the AREOR by May 15 of each year 519 (unless a licensing basis exists for a different submittal date for one or both reports). Licensees use these 520 reports to demonstrate compliance with the facilitys technical specifications for the radioactive effluent 521 control program. The reports demonstrate the following:

522 523 effectiveness of effluent controls and measurement of the environmental impact of radioactive 524 materials 525 526 compliance with the design objectives and limiting conditions for operation required to meet the 527 ALARA criteria in 10 CFR Part 50, Appendix I 528 529 relationship between quantities of radioactive material discharged in effluents and resultant 530 radiation dose to individuals 531 532 compliance with the radiation dose limits to members of the public established by the NRC and 533 the U.S. Environmental Protection Agency (EPA) 534 535 compliance with the effluent reporting requirements of 10 CFR 50.36a1 536 537 Consideration of International Standards2 538 539 The International Atomic Energy Agency (IAEA) works with member states and other partners to 540 promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety 541 Requirements and Safety Guides for protecting people and the environment from harmful effects of 542 ionizing radiation. These requirements and guides provide a system of Safety Standards Categories that 543 reflect an international perspective on what constitutes a high level of safety. In developing or updating 544 Regulatory Guides the NRC has considered IAEA Safety Requirements, Safety Guides1 and other 545 relevant reports in order to benefit from the international perspectives, pursuant to the Commissions 546 International Policy Statement (Ref. 41) and NRC Management Directive and Handbook 6.6 (Ref. 42).

547 548 The following IAEA Safety Standards Series are consistent with the basic safety principles considered in 549 developing this Regulatory Guide:

550 551 IAEA General Safety Guide (GSG)-8, Radiation Protection of the Public and the Environment, 552 issued 2018 (Ref. 43) 553 554 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 14 IAEA Specific Safety Guide NS-G-3.2, Dispersion of Radioactive Material in Air and Water 555 and Consideration of Population Distribution in Site Evaluation for Nuclear Power Plants, 556 issued 2002 (Ref. 44) 557 558 IAEA GSG-9, Regulatory Control of Radioactive Discharges to the Environment, issued 2018 559 (Ref. 45) 560 561 IAEA GSG RS-G-1.8, Environmental and Source Monitoring for Purposes of Radiation 562 Protection, issued 2005 (Ref. 46) 563 564 IAEA Nuclear Energy Series NP-T-3.16, Accident Monitoring Systems for Nuclear Power 565 Plants, issued 2015 (Ref. 47) 566 567 IAEA-TECDOC-482, Prevention and Mitigation of Groundwater Contamination from 568 Radioactive Releases, Vienna, Austria, issued 1988 (Ref. 48) 569 570 IAEA Safety Guide No. WS-G-3.1, Remediation Process for Areas Affected by Past Activities 571 and Accidents, Vienna, Austria, issued 2007 (Ref. 49) 572 573 IAEA, Management of Waste Containing Tritium and Carbon-14, Technical Report Series 574 Number 421, Vienna, Austria, issued 2004 (Ref. 50) 575 576

RG 1.21, Rev. 3, Page 15 C. STAFF REGULATORY GUIDANCE 577 578

1.

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

591 592 1.2 Release Points for Effluent Monitoring 593 594 The ODCM (or equivalent), as required by technical specifications, should identify the facilitys 595 significant release points (see definition in the glossary) used to quantify liquid and gaseous effluents 596 discharged to the unrestricted area. For those release points containing contributions from two or more 597 inputs (or systems), it is preferable to monitor each major input (or system) individually to avoid dilution 598 effects, which may impede or prevent radionuclide identification. NUREG-1301 and NUREG-1302 599 contain detailed guidance for the content and format of a licensees ODCM. For purposes of effluent and 600 direct radiation monitoring, the ODCM should list and describe the following:

601 602

1.

significant release points (see definition in Section 1.3 and in the glossary), which include stacks, 603 vents, and liquid radioactive waste discharge points, among others; 604 605

2.

less-significant release points (see definition in Section 1.4 and in the glossary) that are not 606 normally classified as one of the significant release points but could become a significant release 607 point based on expected operational occurrences (e.g., primary to secondary leakage for PWRs or 608 failed fuel)3; 609 610

3.

the site environs map, which should show each of the following:

611 612

a.

significant release points, 613 614

b.

boundaries of the restricted area and the controlled area4 (in accordance with 615 10 CFR Part 20 definitions),

616 617 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 16

c.

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

d.

boundary of the unrestricted area for gaseous effluents (e.g., the site boundary).

621 622

4.

dose calculation methodologies for exposure pathways and routes of exposure that are identified 623 in RG 1.109, if applicable; and 624 625

5.

dose calculation methodologies for direct radiation if necessary (e.g., when assessing direct 626 radiation from the facility)6.

627 628 1.3 Monitoring a Significant Release Point 629 630 A significant release point is any location from which radioactive material is released that 631 contributes greater than 1 percent of the activity discharged from all the release points for a particular 632 type of effluent considered. RG 1.109 lists the three types of effluent as (1) liquid effluents, (2) noble 633 gases released to the atmosphere, and (3) all other radionuclides discharged to the atmosphere.

634 635 The ODCM should list significant release points. Significant release points should be monitored 636 in accordance with the ODCM. If a new significant release point is identified and is not listed in the 637 ODCM, licensees should (1) establish an appropriate sampling interval (e.g., in site-specific procedures) 638 and (2) update the ODCM within a reasonable timeframe (e.g., annually). Releases from a significant 639 release point should be assessed based on an appropriate combination of actual sample analysis results, 640 radiation monitor responses, flow rate indications, tank level indications, and system pressure indications 641 as necessary to ensure that the amount of radioactive material released, and the corresponding doses, are 642 not substantially underestimated (see 10 CFR Part 50, Appendix I, Section III, Implementation). If 643 activity is detected when monitoring a significant release point, the radionuclides detected should be 644 reported in the effluent totals (including those with half-lives less than 8 days) in the ARERR (i.e., in 645 Table A-1 or Table A-2), provided that the amount discharged is significant to the three-digit exponential 646 format required for the ARERR.

647 648 1.4 Monitoring a Less-Significant Release Point 649 650 NUREG-1301 and NUREG-1302 provide tables designating sampling and analysis frequencies 651 for release points. Historically, these tables, together with the guidance from RG 1.21, Revision 1, issued 652 June 1974 (Ref. 51) or RG 1.21, Revision 2, issued June 2009 (Ref. 52) provide sampling and analysis 653 frequencies. Licensees may continue to use the guidance from NUREG-1301 or NUREG-1302 and/or 654 Revision 1 or Revision 2 of RG 1.21 in accordance with their ODCMs. This method of assigning sample 655 frequencies is simple to implement but, in certain cases, may entail an inappropriately large number of 656 samples for less-significant release points with noor extremely lowimpact on the parameters reported 657 in the ARERR. As a result, for less-significant release points, licensees may evaluate and assign more 658 appropriate sampling frequencies. If a licensee wishes to deviate from the NUREG-1301 and NUREG-659 1302 sampling frequencies, the licensees evaluation must show that the changes (i.e., deviations from 660 NUREG-1301 and NUREG-1302) maintain the levels of radioactive effluent control as stated in the 661 technical specifications required by 10 CFR 20.1302; 40 CFR Part 190; 10 CFR 50.36a; and 662 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 17 10 CFR Part 50, Appendix I, and do not adversely impact the accuracy or reliability of effluent, dose, or 663 setpoint calculations, and should be maintained in site documentation. Regardless of the surveillance 664 frequencies, if activity is detected when monitoring a less-significant release point, the licensee must (in 665 accordance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section III.A.1) report the cumulative 666 activity in the effluent totals (i.e., in Table A-1 or Table A-2) in the ARERR (provided that the amount 667 discharged is significant to the three-digit exponential format required for the ARERR).

668 669 Site documentation should identify less-significant release points, to the extent reasonable, but it 670 is not necessary to list all possible release points in site documentation. Releases from a less-significant 671 release point may be assessed (see Section 5.1) to the extent reasonable using assumptions and bounding 672 calculations (in lieu of, or in addition to, sampling and analysis). When plant conditions change and such 673 changes may reasonably affect the status of a less-significant release point (e.g., significant change in 674 primary-to-secondary leakage in PWRs or substantial cross contamination between systems), the licensee 675 should sample and analyze the affected less-significant release points. These sample results should be 676 evaluated to (1) confirm the continued validity of the bounding calculations (if used) with regard to 677 effluent accountability and (2) determine the impact (if any) on effluent accountability. The guidance in 678 this RG on monitoring less-significant release points for purposes of accountability (through the ARERR) 679 does not replace, supersede, or otherwise modify any responsibility for monitoring systems normally not 680 contaminated, as outlined in NRC Inspection and Enforcement Bulletin 80-10, Contamination of 681 Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity 682 to Environment, issued May 1980 (Ref. 53). A thoroughly designed and documented evaluation of a 683 less-significant release point could also assist in the evaluation and characterization of abnormal releases 684 and abnormal discharges (see Section 1.11 below).

685 686 1.5 Monitoring Leaks and Spills 687 688 An area where an unplanned release occurred in the onsite environs (e.g., a leak or spill) should 689 be identified as an impacted area, as defined in 10 CFR 50.2, Definitions, for decommissioning 690 purposes, and in accordance with NUREG-1757. A leak or spill should be assessed to obtain the 691 necessary information for the ARERR, as specified in Section 8.5.1 of this RG (see glossary).

692 693 Leaks or spills to the ground and/or subsurface will be diluted on contact with soil and water in 694 the environment; therefore, samples of the undiluted liquid (from the source of the leak or spill) and 695 samples of the affected soil (or surface water or subsurface groundwater) should be analyzed as soon as 696 practical. In some instances, sampling, particularly soil sampling, may not be practical if the leak 697 occurred in inaccessible areas or if there are extenuating considerations. In this respect, groundwater 698 monitoring may be used as a surrogate for soil sampling. If sampling is not practical, the 699 10 CFR 50.75(g) records should describe why sampling was not conducted (e.g., the area was 700 inaccessible or there were safety considerations). The licensee should ensure that the location and 701 estimated volume of the leak or spill are recorded to identify the extent of the impacted area and predicted 702 size or extent of the contaminant plume, both horizontally and vertically. If a spill is promptly and fully 703 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 704 residual radioactivity remaining in the soil or groundwater (see paragraph below), for purposes of 705 reporting discharges in the ARERR, there was no liquid discharge to the unrestricted area, and the spill 706 need not be reported in the ARERR. However, in accordance with 10 CFR 50.75(g), the 707 decommissioning file should be updated to include a description of the leak or spill event. Licensees 708 should review the decommissioning files before generating the ARERR to ensure that the ARERR 709 includes the necessary information on leaks and spills.

710 711 When evaluating areas that have been remediated, the licensee should survey for residual 712 radioactivity. There may be times when the licensee wants to verify that an area contains no residual 713

RG 1.21, Rev. 3, Page 18 radioactivity. There is existing regulatory guidance and information on analytical detection capabilities.

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

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

720 721 This RG provides guidance on information that licensees should provide in the ARERR. In that 722 context, when leaks and spills of radioactive material are identified, prompt response and timely actions 723 should be taken to the extent reasonable to (1) evaluate onsite radiological conditions and (2) ensure 724 proper reporting of materials discharged off site. To realize these two goals, it may be necessary to 725 isolate the leak or spill at the source, prevent the spread of the leak or spill, and remediate the affected 726 area (if the licensee deems remediation to be reasonable and necessary).

727 728 For leaks and spills involving the discharge of radioactive material to an unrestricted area, 729 licensees should follow RG 4.25 or equivalent methods to assess the amount of material discharged to the 730 unrestricted area. The potential dose to members of the public from the leak or spill should be evaluated 731 using realistic or bounding exposure scenarios. Attachment 6 to SECY-03-0069, Results of the License 732 Termination Rule Analysis, dated May 23, 2003 (Ref. 54), provides more information on the use of 733 realistic scenarios.

734 735 For leaks and spills, licensees should perform surveys that are reasonable to evaluate the potential 736 radiological hazard (as described in 10 CFR 20.1501). As a result, for leaks and spills, licensees may 737 choose to use bounding assessments to estimate the potential hazard. For example, if a leak occurs on site 738 and radioactive material is released at or below the ground surface, the licensee may choose to assess the 739 potential hazard by estimating a conservatively large (e.g., bounding) volume of water as part of an 740 assumed exposure pathway analysis (e.g., drinking water). Such assumptions would allow the licensee to 741 assess the potential hazard to a hypothetical individual member of the public. A hazard assessment of this 742 sort would be appropriate for inclusion in the supplemental information section of the ARERR. If there is 743 no real exposure pathway to a member of the public, the licensee should indicate that the hazard 744 assessment is a bounding estimate of the dose to a hypothetical individual member of the public, and no 745 real individual member of the public received an actual exposure.

746 747 If licensees choose to notify local authorities of spills or leaks (e.g., because of local ordinances 748 or local and State government agreements), the licensee should review the reporting requirements of 749 10 CFR 50.72(b)(xi) and information in NUREG-1022, Event Reporting Guidelines: 10 CFR 50.72 and 750 50.73, issued October 2000 (Ref. 55), for applicability. In such situations, licensees should ensure 751 effective communication, using NUREG/BR-0308, Effective Risk Communication, issued June 2004 752 (Ref. 56), especially when ensuring that the risk is described in the appropriate context. In general, 753 licensees should notify the NRC when significant public concern is raised, in accordance with 754 10 CFR 50.72(b)(xi).

755 756 Although the licensee may choose to use its problem identification and resolution program 757 (corrective action program) to document the evaluation of the spill or leak, appropriate documentation 758 should be placed in, or cross referenced to, the decommissioning files, as required by 10 CFR 50.75(g).

759 760 Although prompt remediation is not a requirement (Ref. 57), remediation should be evaluated and 761 implemented, as appropriate, based on licensee evaluations and risk-informed decisionmaking. The 762 Electric Power Research Institute (EPRI) Report 1021104 Groundwater and Soil Remediation 763 Guidelines for Nuclear Power Plants, proprietary report issued December 2010 (Ref. 58) and EPRI 764

RG 1.21, Rev. 3, Page 19 Report 1023464, Groundwater and Soil Remediation Guidelines for Nuclear Power Plants, (Public 765 Edition) Final Report, July 2011 (Ref. 59) may be useful in performing remediation evaluations.

766 767 Evaluation factors should include (1) the location and accessibility, (2) the concentrations of 768 radionuclides and extent of the residual radioactivity, (3) the efficacy of monitored natural attenuation, 769 (4) the volume of the release, (5) the mobility of the radionuclides, (6) the depth of the water table, and 770 (7) whether significant residual radioactivity (see glossary) is expected at the time of 771 decommissioning. Since the contaminants, concentrations, and extent of contamination are expected to 772 vary over time or plant life (either increase based on anticipated future leaks and spills or decrease based 773 on remediation or monitored natural attenuation), no one set of numerical values defines significant 774 residual radioactivity. However, licensees may make remediation decisions based on their expectations 775 of their ability to meet the decommissioning criteria of 10 CFR 20.1402 at the anticipated time of 776 decommissioning.

777 778 Information that may be useful in this risk-informed decision making includes (1) NUREG-1757, 779 Volume 1, Appendix H, EPA/NRC Memorandum of Understanding, (2) NUREG-1757, Volume 2, 780 Table H.1, Acceptable License Termination Screening Values of Common Radionuclides for 781 Building-Surface Contamination, and (3) the authorized derived concentration guideline levels for 782 decommissioned nuclear power plants. For a more detailed analysis, licensees may use the computer 783 codes described in NUREG/CR-6676, Probabilistic Dose Analysis Parameter Distributions Developed 784 for RESRAD and RESRAD-BUILD Codes, issued July 2000 (Ref. 60); NUREG/CR-6692, 785 Probabilistic Modules for the RESRAD and RESRAD-BUILD Computer Codes, issued November 786 2000 (Ref. 61); NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 787 3.0 Computer Codes, issued December 2000 (Ref. 62); and NUREG/CR-7267, Default Parameter 788 Values and Distributions in RESRAD-ONSITE V7.2, RESRAD-BUILD V3.5 and RESRAD-OFFSITE 789 4.0 (Ref. 63).

790 791 1.6 Monitoring Continuous Releases 792 793 For continuous releases, gross radioactivity measurements are often the only practical means of 794 continuous monitoring. These gross radioactivity measurements are typically used to actuate alarms and 795 terminate (trip) effluent releases; by themselves, such measurements are generally not acceptable for 796 demonstrating compliance with effluent discharge limits.

797 798 The use of continuously indicating radiation monitoring system results may be combined with 799 sample analyses to more fully characterize and quantify a discharge. This technique may have particular 800 applicability when (1) a short-term, rapid upscale indication of a process radiation monitor occurs during 801 a release or (2) when there is a desire to verify whether a preliminary grab sample is representative. In 802 these instances, the licensee should ensure that the radiation monitor responses (i.e., the radiation monitor 803 efficiencies) for various radionuclides are well characterized.

804 805 Grab samples should be collected at scheduled frequencies in accordance with the ODCM (see 806 NUREG-1301 and NUREG-1302 or as approved in GL 89-01 submittals) to quantify specific 807 radionuclide concentrations and release rates. The frequency of sample collection and radionuclide 808 analyses should be based on the degree of variance in (1) the magnitude of the discharge and (2) the 809 relative radionuclide composition from an established norm. If the magnitude of the discharge and the 810 relative nuclide composition of a continuous release vary significantly over the course of the discharge 811 period, a combination of grab samples and continuous monitor readings can assist in accurately 812 estimating the discharge. Continuous monitoring data (e.g., chart recorder data), as well as grab sample 813 data, should be reviewed periodically and used to identify this variance from the established norm.

814 Periodic evaluations should be made between gross radioactivity measurements and grab sample analyses 815

RG 1.21, Rev. 3, Page 20 of specific radionuclides. These evaluations should be used to verify (or modify) the conversion factors 816 that correlate radiation monitor readings and concentrations of radionuclides in effluents.

817 818 NUREG-1301 and NUREG-1302 provide guidance on the Radiological Environmental 819 Monitoring Program. Table 3.12-1 therein provides guidance on implementing the environmental 820 monitoring program, including an I-131 sampling and analysis on each composite of drinking water.

821 822 If a drinking water exposure pathway exists, a prospective dose evaluation should be performed 823 based on I-131 in effluent discharges to determine the maximum likely annual I-131 thyroid dose to a 824 person in any age group. The purpose of the prospective dose evaluation is to determine the 825 environmental sampling and analysis requirements. Note: Freshwater fish ingestion is not included in 826 the prospective dose evaluation of I-131 from the drinking water route of exposure.

827 828 If the likely dose from I-131 is greater than 1 mrem per year, a composite drinking water sample 829 should be collected over a 2-week period and an I-131 analysis performed with an LLD of 1 pCi/liter. If 830 the likely dose from I-131 is less than or equal to 1 mrem per year, a monthly composite sample should be 831 collected and an I-131 analysis performed with an LLD of 15 pCi/liter.

832 833 In addition, Standard Technical Specifications require determination of the projected dose 834 contributions from radioactive effluents at least every 31 days, and determination of the cumulative dose 835 contributions for the current calendar quarter and current calendar year.

836 837 1.7 Monitoring Batch Releases 838 839 For batch releases, measurements should be performed to identify principal radionuclides before 840 a release. If an analysis of specific hard-to-detect radionuclides (such as strontium-89/90, Ni-63 and 841 iron-55 in liquid releases) cannot be done before the batch release (see NUREG-1301 and NUREG-1302),

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

847 848 Continuously indicating radiation monitoring system results may be combined with sample 849 analyses to more fully characterize and quantify a discharge. This technique may have particular 850 applicability when (1) a short-term, rapid upscale indication of a process radiation monitor occurs during 851 a discharge or (2) when there is a desire to verify whether a preliminary grab sample is representative. In 852 these instances, the licensee should ensure that radiation monitor responses (i.e., the radiation monitor 853 efficiencies) for various radionuclides are well characterized.

854 855 1.8 Principal Radionuclides for Effluent Monitoring 856 857 This RG introduces the term principal radionuclide in a risk informed context. A licensee may 858 evaluate the list of principal radionuclides for use at a particular site. The principal radionuclides maybe 859 determined based on their relative contribution to either (1) the public dose compared to the 860 10 CFR Part 50, Appendix I, design objective doses, or (2) the amount of activity discharged compared to 861 other site radionuclides in the type of effluent being considered. Under this concept, radionuclides that 862 have either a significant activity or a significant dose contribution should be monitored in accordance 863 with a predetermined and appropriate analytical sensitivity level (LLD) outlined in a licensees ODCM.

864 This implementation of principal radionuclides ensures that the ARERR appropriately includes both the 865 (1) radionuclides that are present in relatively large amounts but that contribute very little to dose and 866

RG 1.21, Rev. 3, Page 21 (2) radionuclides that are present in very small amounts but that have a relatively high contribution to 867 dose.

868 869 If a risk-informed approach is used, principal radionuclides should be determined based on an 870 evaluation over a time period that includes a refueling outage (e.g., one fuel cycle). A periodic 871 reevaluation should be performed to determine whether the radionuclide mix has changed and to identify 872 new principal radionuclides.

873 874 If a risk-informed approach is applied to the determination of principal radionuclides,7 the ODCM 875 becomes the controlling document and specifies the list of principal radionuclides. If adopting this 876 method, the licensee should update the ODCM with the list of principal radionuclides within 1 year of 877 their identification. Licensees are allowed to revise the ODCM in accordance with the ODCM change 878 process, as described in the plants technical specifications (which includes documented evaluations of 879 such changes).

880 881 If adopting a risk-informed approach, a radionuclide is considered a principal radionuclide if it 882 contributes either (1) greater than 1 percent of the 10 CFR Part 50, Appendix I, design objective dose for 883 all radionuclides in the type of effluent being considered or (2) greater than 1 percent of the activity of all 884 radionuclides in the type of effluent being considered. RG 1.109 lists the three types of effluent as 885 (1) liquid effluents, (2) noble gases released to the atmosphere, and (3) all other radionuclides released to 886 the atmosphere. In this context, the term principal radionuclide has special significance for the required 887 sensitivity levels (e.g., LLDs) for an analysis. The LLDs specified in NUREG-1301 and NUREG-1302 888 may be used, or LLDs may be determined based on the other methodologies (e.g., as outlined in 889 NUREG-1576). Once principal radionuclides are identified, they should be monitored in accordance with 890 the sensitivity levels (e.g., LLDs) listed in the ODCM.

891 892 During analysis of samples, licensees should apply the appropriate analytical sensitivities to 893 ensure adequate surveys are conducted. NUREG-1301 and NUREG-1302 provide a list of principal 894 gamma emitters for operating reactors for which an LLD control applies. Historically, this list and 895 guidance from Revision 1 or Revision 2 provided the appropriate sensitivity levels for an analysis.

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

902 903 Although the LLD list from NUREG-1301 and NUREG-1302 may be used to determine principal 904 radionuclides, in reality, the principal radionuclides at a site will depend on site-specific factors, such as 905 (1) the operating status of the facility (e.g., operating or in decommissioning), (2) the amount of failed 906 fuel, (3) the extent of system leakage, (4) the sophistication of radioactive waste processing equipment, 907 and (5) the level of expertise in operating radioactive waste processing systems. Since the principal 908 radionuclides will vary from site to site, licensees that wish to deviate from the historical method of 909 determining principal radionuclides (as described above) may adopt a risk-informed approach to identify 910 principal radionuclides (and the associated sensitivity levels) at a site.

911 912 For radionuclides that are not identified as principal radionuclides, licensees may use their 913 discretion with the sensitivity of analysis, provided the licensees determine that the changes maintain the 914 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 22 levels of radioactive effluent controls required by the regulations in 10 CFR 20.1302; 40 CFR Part 190; 915 10 CFR 50.36a; and 10 CFR Part 50, Appendix I, and do not adversely impact the accuracy or reliability 916 of effluent, dose, or setpoint calculations. If licensees change their analytical sensitivities from those in 917 their ODCM or equivalent, they must document the basis for the deviations. For example, DQOs and 918 other concepts from RG 4.15 may be useful for determining risk-informed sensitivity levels for an 919 analytical method.

920 921 The risk-informed concept of principal radionuclides does not reduce the requirement for 922 reporting radionuclides detected in effluents. In addition to principal radionuclides, other radionuclides 923 detected during routine monitoring of release points should be reported in the radioactive effluent release 924 report and included in dose assessments to members of the public, consistent with site-specific technical 925 specifications.

926 927 1.9 Carbon-14 928 929 Carbon (C)-14 is a naturally occurring isotope of carbon. Nuclear weapons testing in the 1950s 930 and 1960s significantly increased the amount of C-14 in the atmosphere. Commercial nuclear reactors 931 also produce C-14 but in much lower amounts than those produced naturally or from weapons testing.

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

935 936 Regulations in 10 CFR 50.36a require that operating procedures be developed for the control of 937 effluents and that quantities of principal radionuclides be reported. The radioactive effluents from 938 commercial nuclear power plants overtime has decreased to the point that C-14 is likely to have become a 939 principal radionuclide (as defined in this document) in gaseous effluents. Therefore, licensees must 940 evaluate whether C-14 is a principal radionuclide for gaseous releases from their facility. Because the 941 dose contribution of C-14 from liquid radioactive waste is much less than that contributed by gaseous 942 radioactive waste, an evaluation of C-14 in liquid radioactive waste is not required.

943 944 The quantity of C-14 discharged can be estimated by use of a normalized C-14 source term and 945 scaling factors based on power generation or estimated by use of the NUREG-0016 (GALE-BWR) and 946 NUREG (GALE-PWR) computer codes. The National Council on Radiation Protection and 947 Measurements Report No. 81, Carbon-14 in the Environment, (Ref. 64) also provides information about 948 the magnitude of C-14 in typical effluents from commercial nuclear power plants. These documents 949 estimate that nominal annual releases of C-14 in gaseous effluents are approximately from 5 to 7.3 curies 950 from PWRs and from 8 to 9.5 curies from BWRs.

951 952 The quantity of C-14 generated in BWR and PWR cores can also be estimated by a calculational 953 method provided by the EPRI Report No. 1021106, Estimation of Carbon-14 in Nuclear Power Plant 954 Gaseous Effluents, issued December 2010 (Ref. 65) and EPRI Report No. 1024827 "Carbon-14 Dose 955 Calculation Methods at Nuclear Power Plants," issued April 2012, (Ref. 66). If estimating C-14 based on 956 scaling factors and fission rates, a precise and detailed evaluation of C-14 is not necessary. It is not 957 necessary to calculate uncertainties for C-14 or to include C-14 uncertainty in any subsequent calculation 958 of overall uncertainty.

959 960 Since the NRC published RG 1.21, Revision 1, in 1974, the analytical methods for determining 961 C-14 have improved. Because the production of C-14 is expected to be relatively constant at a particular 962 site, if sampling is performed for C-14 (instead of estimating C-14 discharges based on calculations), the 963 sampling frequency may be adjusted to that interval that allows adequate measurement and reporting of 964 effluents.

965

RG 1.21, Rev. 3, Page 23 966 1.10 Return/Reuse of Previously Discharged Radioactive Effluents 967 968 Radioactive material properly released in gaseous or liquid effluents to the unrestricted area 969 (excluding solid materials or soil) is not considered licensed material when returned to the facility as long 970 as the concentration of radioactive material does not exceed 10 CFR Part 30, Rules of General 971 Applicability to Domestic Licensing of Byproduct Material, exempt concentration limits (otherwise a 972 general or specific license is required). The water containing radioactive material returned from the 973 environment can be used by the licensee and returned to the unrestricted area without being considered a 974 new radioactive material effluent release. The basis for this determination is that the licensee has already 975 accounted for this radioactive material when the effluent was originally discharged, provided that the 976 subsequent use, possession, or release does not introduce a new significant dose pathway to a member of 977 the public, as explained below.

978 979 Licensees are responsible for evaluating any new significant exposure pathway and the resultant 980 radiological hazards associated with the return of radioactive material to the operating facility and its 981 subsequent discharge to the environment. For purposes of estimating dose during operations or 982 decommissioning, a new significant exposure pathway is any pathway that contributes dose that exceeds 983 10% of the dose criteria in 10 CFR 50 Appendix I, Section II (such that the dose from a new exposure 984 pathway is unlikely to be substantially underestimated). Bounding dose assessments as described in 985 Section 5.1 of this RG may be used in evaluating any new significant exposure pathway. Furthermore, 986 before returning radioactive materials to the environment, licensees must demonstrate that these 987 radioactive materials were previously disposed of in accordance with 10 CFR 20.2001(a)(3), or that the 988 material is naturally occurring background radiation. Radioactive material previously not accounted for as 989 an effluent that is entrained with returned/re-used water must be considered a new effluent disposal per 10 990 CFR 20.2001. See RIS 2008-03 for further details.

991 992 1.11 Abnormal Releases and Abnormal Discharges 993 994 In RG 1.21, Revision 1, the terms release and discharge were synonymous. In RG 1.21, 995 Revision 2 and 3, the term release describes an effluent from the plant (regardless of where the effluent 996 is located), and the term discharge describes an effluent that enters the unrestricted area. Although the 997 term release includes effluents to either (1) the onsite environs or (2) the unrestricted area, this RG 998 generally reserves use of the term release for the release of an effluent from the power plant into the 999 onsite environs. The onsite environs in this context encompass locations outside of nuclear power plant 1000 systems, structures, and components, as described in the final safety analysis report or ODCM. This is a 1001 change in terminology with respect to the definition of abnormal release in RG 1.21, Revision 1, which 1002 defined abnormal releases to be from the site boundary.

1003 1004 An abnormal release (see glossary) is an unplanned or uncontrolled release of licensed 1005 radioactive material into the onsite environs. Abnormal releases may be categorized as either batch or 1006 continuous, depending on the circumstances. By contrast, an abnormal discharge (see glossary) is an 1007 unplanned or uncontrolled discharge of licensed radioactive material to the unrestricted area. Abnormal 1008 discharges may also be categorized as either batch or continuous, depending on the circumstances. The 1009 distinction between the terms abnormal release and abnormal discharge is important for describing 1010 the staff position for measuring, evaluating, and reporting releases and discharges, especially where leaks 1011 and spills are involved.

1012 1013 That portion of an abnormal release discharged to the unrestricted area is reported as an abnormal 1014 discharge in the year in which the discharge to the unrestricted area occurred. The portion of an abnormal 1015

RG 1.21, Rev. 3, Page 24 release that remains onsite is considered residual radioactivity (see 10 CFR Part 20) and is documented in 1016 accordance with 10 CFR 50.75(g).

1017 1018 Low-level radioactive system leakage resulting from minor equipment failures and component 1019 aging (wear and tear) may be expected to occur as an anticipated part of the plant operation. If such 1020 leakage is captured by, or directed to, a system designed to accept and handle radioactive material, 1021 including the subsequent planned and controlled discharge of the radioactive material (e.g., as described 1022 in the final safety analysis report or ODCM), that evolution is not considered an abnormal release.

1023 Normal system leakage captured by effluent ventilation control systems or sumps is not an abnormal 1024 release (provided that, before discharge of the radioactive material, the discharge is planned and 1025 controlled). (See also the definitions of unplanned release and uncontrolled release in the glossary.)

1026 1027 In certain circumstances, some subjectivity may be associated with the definitions of unplanned 1028 release and uncontrolled release. In these situations, additional circumstances should be considered to 1029 determine whether an abnormal release occurred. A well-designed and documented evaluation of a 1030 release point can include an evaluation of the potential for an unplanned or uncontrolled release. The 1031 evaluation can establish bounding criteria that establish a threshold for an abnormal release based on 1032 planning and control. Generally, releases that may reasonably be categorized as both unplanned and 1033 uncontrolled should be considered abnormal releases.

1034 1035 For example, consider an underground pipe that carries radioactive liquid to an outside storage 1036 tank. If this pipe develops a leak, and licensed radioactive material escapes into the surrounding soil, it is 1037 considered an abnormal release if some portion or all of the radioactive material remains onsite. This 1038 type of leak should be reported as an abnormal release in the next ARERR. If the licensee predicts 1039 (e.g., based on site conceptual model and subsequent groundwater monitoring results) that the radioactive 1040 material will enter the unrestricted area in 2 years, the resulting radioactive discharge (that would occur 1041 2 years hence) will be considered an abnormal discharge. Therefore, the resulting radioactive discharge 1042 should be reported along with other data for the affected calendar year in a future ARERR (i.e., in this 1043 example, 3 years later). Both releases and discharges (either routine or abnormal) should be reported on a 1044 calendar-year basis for the year in which the release or discharge occurred.

1045 1046 Consider another example involving a volume of radioactive gas from the containment 1047 atmosphere that escapes the equipment hatch during a refueling outage (especially during the time 1048 interval when the containment purge exhaust fans are off). This would generally not be considered an 1049 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 />),

1050 (2) the containment activity (gas, particulate, tritium, and iodine) was preplanned, known, and very low 1051 (e.g., such that a bounding estimate of the radioactive material discharged indicated there would be no 1052 measurable impact relative to typical discharges), (3) the containment activity was monitored (e.g., by 1053 sampling or radiation monitoring equipment), and (4) an evaluation was completed to identify a 1054 preplanned limiting (or trigger) level of activity that would initiate remedial or mitigating action 1055 (e.g., close the equipment hatch to control gases escaping containment). In this example, the actions 1056 taken (i.e., preplanning and monitoring) before and during the evolution are sufficient to establish control 1057 of this discharge. As a result, this type of evolution should not be categorized as an abnormal discharge.

1058 1059

2.

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

RG 1.21, Rev. 3, Page 25 samples should be obtained as needed to characterize abnormal releases, abnormal discharges, or other 1067 significant operational evolutions. Samples should be representative of the overall effluent in the bulk 1068 stream, collection tank, or container. Licensees should ensure that representative samples were obtained 1069 from well-mixed streams or volumes of effluent at sampling points, using proper equipment and sampling 1070 procedures.

1071 1072 2.2 Sampling Liquid Radioactive Waste 1073 1074 Before sampling, large volumes of liquid waste should be mixed to ensure that sediments or 1075 particulate solids are distributed uniformly in the waste mixture. For example, a large tank may be mixed 1076 using a sparger system or recirculated three or more volumes to ensure that a representative sample can be 1077 obtained, as recommended by American Society for Testing and Materials (ASTM) D 3370 - 18, 1078 Standard Practices for Sampling Water from Flowing Process Streams (Ref. 67). If tank-mixing 1079 practices deviate from industry standards (i.e., those for recirculation or otherwise), the licensee should 1080 provide a technical evaluation or other justification. Sample points should be located where there is a 1081 minimum of disturbance of flow caused by fittings and other physical characteristics of the equipment 1082 and components. Sample nozzles should be inserted into the flow or liquid volume to ensure sampling of 1083 the bulk volume of pipes and tanks. Sample lines should be flushed for a sufficient period of time before 1084 sample extraction to remove sediment deposits and air and gas pockets. Generally, three sample line 1085 volumes should be purged as recommended by ASTM D3370 - 18, before withdrawing a sample, unless a 1086 technical evaluation or other justification is provided. A series of samples should be taken periodically 1087 during the interval of discharge to determine whether any differences exist as a function of time and to 1088 ensure that individual samples are indeed representative of the effluent mixture. In some instances, this 1089 may be accomplished by collecting one or more samples (either by grab or composite sampler) during 1090 the discharge and comparing with one or more samples taken before the discharge. If a series of samples 1091 is collected, these samples can be used to assess the amount of measurement uncertainty in obtaining 1092 representative samples.

1093 1094 2.3 Sampling Gaseous Radioactive Waste 1095 1096 Although all licensees may not be committed to RG 4.15, ANSI N42.18-2004, Specification and 1097 Performance of On-Site Instrumentation for Continuously Monitoring Radioactivity in Effluents 1098 (Ref. 68), ANSI N42.54-2018, Instrumentation and Systems for Monitoring Radioactivity (Ref. 69),

1099 and ANSI/Health Physics Society (HPS) N13.1-2011, Sampling and Monitoring Releases of Airborne 1100 Radioactive Substances from the Stacks and Ducts of Nuclear Facilities (Ref. 70), these documents 1101 provide general principles for designing and conducting monitoring programs for airborne effluents. The 1102 cited references also contain recommendations for obtaining valid samples of airborne radioactive 1103 material in effluents and the guidelines for sampling from ducts and stacks. Licensees should use the 1104 appropriate licensing documents to evaluate the validity of representative samples (e.g., evaluate the 1105 potential for inaccurate sampling of gaseous effluents that may bypass a particulate filter and collect on an 1106 iodine collection cartridge) and to identify any inaccurate sample analyses configurations or counting 1107 geometries.

1108 1109 2.4 Sampling Bias 1110 1111 Sampling and storage techniques that could bias quantitative results for effluent measurements 1112 should be evaluated and corrections applied as necessary. These biases include inaccurate measurement 1113 of sample volumes resulting from pressure drops in long sample lines and loss of particulates or iodine in 1114 sample lines resulting from deposition or plate-out. Samplers for gaseous waste should be evaluated for 1115 particulate deposition using ANSI/HPS N13.1-1999 or equivalent.

1116 1117

RG 1.21, Rev. 3, Page 26 2.5 Composite Sampling 1118 1119 Composite samples should be representative of the average quantities and concentrations of 1120 radioactive materials discharged in liquid and gaseous effluents. Composite samples should be collected 1121 in proportion to the effluent flow rate or in proportion to the volume of each batch of effluent discharges.

1122 1123 2.6 Sample Preparation and Preservation 1124 1125 Sample preparation and storage methods should minimize the potential for loss of radioactive 1126 material (i.e., deposition of analyte on walls of the sample container or volatilization of analyte).

1127 Composite sample storage time should be as short as practical to preclude deposition on the storage 1128 container, or sample stabilization should be considered. Before quantitative radionuclide analyses for 1129 liquid effluent composites, licensees should ensure that samples are mixed thoroughly so that the sample 1130 is representative of the material discharged.

1131 1132 Procedures for handling, packaging, and storing samples should ensure that losses of radioactive 1133 materials or other factors causing sample deterioration do not invalidate the analysis. For example, filters 1134 should be stored carefully to prevent loss of radioactive material from the filter paper.

1135 1136 2.7 Short-Lived Radionuclides and Decay Corrections 1137 1138 In the analysis of short-lived radionuclides (e.g., short-lived noble gases), measurements should 1139 generally be made as soon as practical after collection to minimize loss by radioactive decay. In other 1140 cases, when needed to improve the detection of the longer-lived radionuclides, time should be allowed for 1141 the decay of short-lived, interfering radionuclides.

1142 1143 Some special considerations may be applicable when measuring short-lived radionuclides. In 1144 general, sample collection (or analysis frequencies) should take into account the half-lives of the 1145 radionuclides being measured. This may have special applicability for continuous samples or composite 1146 samples. It is generally best to select a compositing interval (and analysis frequency) appropriate for the 1147 effluent (radionuclide) being analyzed. In cases where the compositing interval is selected appropriately, 1148 analytical bias is minimized. One way to avoid analytical bias is to decrease the composite sampling 1149 interval (and analysis frequency).

1150 1151 To minimize bias in measurements, it may be necessary to decay correct analysis results for 1152 short-lived radionuclides. Licensees should be cognizant of those situations in which analytical bias may 1153 be introduced when analyzing short-lived radionuclides and should select appropriate methods to 1154 minimize such bias.

1155 1156

3.

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

RG 1.21, Rev. 3, Page 27 maintain an annual meteorological summary report that provides the joint frequency distributions of wind 1169 direction and wind speed by atmospheric stability class (see RG 1.23, or, if applicable, Safety Guide 23, 1170 Onsite Meteorological Programs, dated February 17, 1972 (Ref. 71)) on site for the life of the plant. In 1171 addition, the licensee should record hourly meteorological data (or shorter-term averages compatible with 1172 the appropriate dispersion models) and make the data available if needed for assessing abnormal gaseous 1173 releases.

1174 1175 3.2 Atmospheric Dispersion (Transport and Diffusion) 1176 1177 Site-specific meteorological data collected should be validated and used to generate gaseous 1178 effluent dispersion factors (/Q) and deposition factors (D/Q), in accordance with RG 1.111. The use of 1179 long-term annual-average meteorological conditions (based on 5 or more years of data) to determine /Q 1180 and D/Q is appropriate for continuous releases and for establishing instantaneous release rate set points.

1181 This practice may also be acceptable for calculating doses from intermittent releases if the releases occur 1182 randomly and with sufficient frequency to justify the use of annual-average meteorological conditions 1183 (see RG 1.111).

1184 1185 Personnel familiar with the equipment and typical site meteorological conditions should review 1186 the meteorological data. Data losses can be minimized by incorporating redundant sensors and 1187 equipment, and by maintaining an adequate inventory of spares, as part of the monitoring program design.

1188 Periodic data evaluation may include, but is not be limited to, promptly identifying and inspecting 1189 equipment failures and time to resolution, reviewing results of performance checks and calibrations, and 1190 confirming that measurements are within appropriate ranges (e.g., occurrence of excessive calm wind 1191 speeds, reasonable diurnal and seasonal variation of wind speed, wind direction, and temperature at each 1192 level and with height).

1193 1194 A change in /Q (and/or D/Q) may not be the only indicator that should be reviewed. A change 1195 in impact location should also be addressed (if not already the case). Such a change could be caused by 1196 (1) an actual change in the meteorological conditions, (2) a physical change in meteorological 1197 instrumentation (i.e., mechanical versus sonic anemometry), (3) a change in data averaging approach 1198 (e.g., scalar versus vector), or (4) any combination of the above.

1199 1200 Invalid data should be removed from the meteorological data file prior to calculating long-term, 1201 annual-average /Q and D/Q values. Records of data invalidation (and if applicable, data substitution) 1202 should also be documented and retained.

1203 1204 The long-term, annual-average /Q and D/Q values should be reevaluated periodically (e.g., every 1205 3-5 years). If the periodic reevaluation indicates the controlling/limiting long-term, annual-average /Q 1206 and D/Q values are substantially nonconservative (e.g., higher by 20-30 percent or more with respect to 1207 historical data), the licensee should ensure that the /Q and D/Q values used in the dose assessment are 1208 revised or that the ARERR addresses why such changes are not deemed necessary. Acceptable reasoning 1209 includes evaluating data anomalies, identification of failures in meteorological sensors, and 1210 documentation that the locality experienced abnormal weather patterns.

1211 1212 3.3 Release Height 1213 1214 The release height affects the dispersion (transport and diffusion) of radioactive materials, 1215 especially for downwash and building wake effects. For facilities with ground-level, mixed-mode, and 1216 elevated releases, an evaluation should be made to determine the proper location of the maximum 1217 exposed individual member of the public. From a dispersion perspective, when determining the 1218 maximum exposure location (submersion and/or deposition), the evaluation should consider the 1219

RG 1.21, Rev. 3, Page 28 magnitude of the release(s) originating as an elevated release and as a ground-level release. For example, 1220 a close-in, downwind location in one sector may have a higher /Q (i.e., less dispersion) for a 1221 ground-level release, whereas the majority of the source term may be originating as an elevated release, 1222 causing a higher concentration () at a more distant location, possibly in a different sector. RG 1.111 1223 contains a more complete discussion of release height.

1224 1225 3.4 Aquatic Dispersion (Surface Waters) 1226 1227 Liquid radioactive effluents may be disposed in accordance with 10 CFR 20.2001 into a variety 1228 of receiving surface water bodies, including nontidal rivers, lakes, reservoirs, settling ponds, cooling 1229 ponds, estuaries, and open coastal waters. This effluent is dispersed by various mechanisms 1230 (i.e., turbulent mixing; stream flow in the water bodies; and internal circulation or flow-through in lakes, 1231 reservoirs, and cooling ponds). Parameters influencing the dispersion patterns and concentrations near a 1232 site include the direction and speed of flow of currents, both natural and plant induced, in the receiving 1233 water; the intensity of turbulent mixing; the size, geometry, and bottom topography of the receiving 1234 water; the location of effluent discharge in relation to the receiving water surface and shoreline; the 1235 amount of recirculation of previously discharged effluent; the characteristics of suspended and bottom 1236 sediments; and sediment sorption properties. RG 1.113 describes calculational models for estimating 1237 aquatic dispersion to surface water bodies. However, the dispersion characteristics may be highly site 1238 dependent, and local characteristics should be considered when performing dispersion modeling and dose 1239 assessments.

1240 1241 3.5 Spills and Leaks to the Ground Surface 1242 1243 Liquid releases onto the land surface are transported and diluted as a function of site-specific 1244 hydrologic features, events, and processes and properties of the effluent. The releases may temporarily 1245 accumulate, pool, or run off to natural or engineered drainage systems. During this process, water may 1246 also be absorbed into the soil (see Section 3.6). RG 1.113 discusses the use of simple models to estimate 1247 transport through surface water bodies and considers water usage effects. Spills or leaks of radioactive 1248 material to the ground surface should initiate characterization of the runoff. At a minimum, the 1249 characterization activities should satisfy (1) the requirements of 10 CFR 50.75(g) and (2) the effluent 1250 reporting requirements of 10 CFR 50.36a, and the guidance described in NUREG-1301 and 1251 NUREG-1302 for planned effluents (e.g., sampling before discharge to unrestricted areas).

1252 Sections 8.5.1, 8.5.2, and 8.5.9 of this RG contain recommendations on the general format for reporting 1253 abnormal releases to onsite areas and abnormal discharges to unrestricted areas.

1254 1255 3.6 Spills and Leaks to Groundwater 1256 1257 Liquid radioactive leaks and spills are sometimes released to onsite groundwater or discharged to 1258 offsite groundwater. Leaks and spills onto the ground surface can be absorbed into the soil. Depending 1259 on the local soil properties and associated liquid flux of the release, some of the material in the leak or 1260 spill may eventually reach the local water table. The dispersion of this material depends on the local 1261 subsurface geology and hydrogeologic characteristics. Liquid releases into the subsurface will be 1262 transported as a function of groundwater flow processes and conditions (e.g., hydraulic gradients, 1263 permeability, porosity, and geochemical processes) and will eventually be released to the unrestricted 1264 area.

1265 1266 A groundwater conceptual site model should be developed to predict the subsurface water flow 1267 parameters to include direction and rate and to be used as the basis for estimating the dispersion of 1268 abnormal releases of liquid effluents into groundwater (see RG 4.1 and RG 4.25). Section 1 of this RG 1269 lists references for use in developing an adequate groundwater conceptual site model.

1270

RG 1.21, Rev. 3, Page 29 1271 Simple analytical models or more rigorous numerical codes (i.e., simulations) may be used to 1272 evaluate subsurface transport following a release. Appropriate use of these models and codes will depend 1273 on the release rate, depth of the release, depth to the local water table, groundwater flow directions, 1274 groundwater flow rates, geochemical conditions, and other geochemical processes (e.g., geochemical 1275 retardation). Additionally, water usage, such as groundwater pumping from wells, may create local 1276 groundwater depression(s) that can alter the natural groundwater flow.

1277 1278 Consistent with 10 CFR 20.1501, a basic site hydrogeological characterization, in advance of 1279 leaks or spills, is helpful for evaluating potential leaks and spills. Sites with significant residual 1280 radioactivity that are likely to exceed the radiological criteria for unrestricted use at the time of 1281 decommissioning (e.g., as described in 10 CFR 20.1402) should perform more extensive evaluation.

1282 Initial assessments should be conducted with relatively simple conceptual site models using scoping 1283 surveys, bounding assumptions, or a combination of both (see RG 4.25 and American National Standards 1284 Institute/American Nuclear Society (ANSI/ANS) 2.17-2009, Evaluation of Subsurface Radionuclide 1285 Transport at Commercial Nuclear Power Production Facilities (Ref. 72). The complexity of the models 1286 should increase as (1) more knowledge is obtained about the system under evaluation (e.g., source of leak, 1287 plume size, concentrations, radionuclides, site characteristics, presence of preferential flow pathways) and 1288 (2) the dose estimates rise above significant residual radioactivity levels (see definition in the glossary).

1289 Industry documents (Refs. 38 and 72) contain details of various industry practices that may be used as 1290 part of a groundwater monitoring program. Sites with low-level spills or leaks generally do not require 1291 extensive site characterization and monitoring.

1292 1293 The following are basic steps in monitoring groundwater contamination:

1294 1295

1.

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

1297 1298

2.

Collect empirical data by one or more of the following (as necessary):

1299 1300

a.

Sample and analyze groundwater from existing monitoring wells.

1301 1302

b.

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

1303 1304

3.

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

1306 1307

4.

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

1309 1310

5.

Use an iterative process and revaluate as needed.

1311 1312 The groundwater monitoring results should be used in the development and testing of a 1313 conceptual site model to predict radionuclide transport in groundwater. The conceptual site model is 1314 generally considered adequate when it predicts the results of monitoring (sometimes called a calibrated 1315 model). Groundwater monitoring results evaluate the validity of the conceptual site model. Following a 1316 leak or spill of licensed (radioactive) material, the conceptual site model may be used in conjunction with 1317 radionuclide transport modeling and groundwater monitoring to comprise a basis for predicting future 1318 effluents from the site. Dispersion and dilution occur over time and in three dimensions.

1319 1320

RG 1.21, Rev. 3, Page 30 When used with a strategic and carefully planned monitoring program, the conceptual site model 1321 can ensure that necessary and reasonable surveys are performed (i.e., limited scoping surveys or more 1322 extensive surveys). Limited scoping surveys can determine if significant residual radioactivity exists and 1323 if there is adequate protection of public health and safety. If the limited scoping surveys identify 1324 significant residual radioactivity, then the extent of the contamination should be further evaluated by 1325 more extensive surveys (e.g., monitoring wells or other evaluations as appropriate). These survey 1326 activities may be direct (i.e., occurring at, or very near, the source of the leak) or indirect (i.e., occurring 1327 at some distance from the source of the leak) depending on the accessibility of the source of the spill or 1328 leak and the mobility of the radionuclides.

1329 1330 For spills or leaks occurring below the soil surface in inaccessible locations, direct scoping and 1331 characterization may not be feasible. In these cases, indirect monitoring techniques (e.g., groundwater 1332 monitoring wells in a down-gradient direction) will satisfy existing regulatory requirements. These 1333 survey activities should, at a minimum, satisfy (1) the requirements of 10 CFR 50.75(g) and (2) the 1334 effluent reporting requirements of 10 CFR 50.36a for groundwater discharges to the unrestricted area. In 1335 general, licensees should describe (report) leaks and spills of radioactive material in the ARERR for the 1336 calendar year the spill or leak occurred. Additionally, licensees should report groundwater monitoring 1337 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 1338 of this RG contain guidance on the general format for reporting abnormal releases to onsite areas and 1339 abnormal discharges to unrestricted areas.

1340 1341 Although licensees may conduct a groundwater monitoring effort for different reasons, for 1342 purposes of this RG, the surveys, characterization activities, conceptual site models, and other 1343 components of any groundwater monitoring effort should be sufficient to do the following:

1344 1345

1.

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

1346 1347

2.

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

1348 1349

3.

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

1351 1352

4.

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

1354 1355

5.

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

1356 1357

4.

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

1370 1371

RG 1.21, Rev. 3, Page 31 4.2 Quality Control Checks 1372 1373 QC checks of laboratory instrumentation should be conducted daily or before use, and 1374 background variations should be monitored at regular intervals to demonstrate that a given instrument is 1375 in working condition and functioning properly. QC records should include results of routine tests and 1376 checks, background data, calibrations, and all routine maintenance and service.

1377 1378 4.3 Surveillance Frequencies 1379 1380 Routine qualitative tests and checks (e.g., channel operational tests, channel checks, or source 1381 checks to demonstrate that a given instrument is in working condition and functioning properly) may be 1382 performed using radioactive sources that are not traceable by the National Institute of Standards and 1383 Technology (NIST). The schedule for source checks, channel checks, channel calibrations, and channel 1384 operational tests should be in accordance with NUREG-1301 and NUREG-1302, unless otherwise 1385 modified after a technical evaluation demonstrates a justifiable change in frequency. A technical 1386 evaluation that revises a surveillance frequency should include consideration of the instruments function 1387 and the consequences of failure and not simply rely on the history of successful surveillances.

1388 1389 4.4 Procedures 1390 1391 Individual written procedures should be used to establish specific methods of calibrating installed 1392 radiological monitoring systems and grab sampling equipment. Written procedures should document 1393 calibration practices used for ancillary equipment and systems (e.g., meteorological equipment, airflow 1394 measuring equipment, in-stack monitoring pitot tubes). Calibration procedures may be compilations of 1395 published standard practices or manufacturers instructions that accompany purchased equipment, or they 1396 may be written in house to include special methods or items of equipment not covered elsewhere.

1397 Calibration procedures should identify the specific equipment or group of instruments to which the 1398 procedures apply.

1399 1400 Written procedures should be used for maintaining counting room instrument accuracy, including 1401 maintenance, storage, and use of radioactive reference standards; instrumentation calibration methods; 1402 and QC activities such as collection, reduction, evaluation, and reporting of QC data as required by the 1403 technical specifications.

1404 1405 4.5 Calibration of Laboratory Equipment and Routine Effluent Radiation Monitors 1406 1407 Calibrations (e.g., of laboratory equipment and continuous radiation monitoring systems used to 1408 quantify radioactive effluents) should be performed using the general principles for calibration of effluent 1409 monitoring instrumentation provided in ANSI N42.18-2004 and ANSI N323C-2009, American National 1410 Standard for Radiation Protection Instrumentation Test and CalibrationAir Monitoring Instruments, 1411 American National Standards Institute (Ref. 74), using radioactive calibration sources traceable to the 1412 NIST. Calibration sources should have the necessary accuracy, stability, and radioactivity levels required 1413 for their intended use. The relationship between concentrations and monitor readings should be 1414 determined. Performance of the monitoring system should be judged on the basis of reproducibility, time 1415 stability, and sensitivity.

1416 1417 Periodic inservice correlations that relate monitor readings to the concentrations, release rates of 1418 radioactive material in the monitored release path, or a combination of both, should be performed when 1419 possible to validate the adequacy of the system. These correlations should be based on the results of 1420 analyses for specific radionuclides in grab samples from the release path.

1421 1422

RG 1.21, Rev. 3, Page 32 The use of NIST-traceable sources combined with mathematical efficiency calibrations may be 1423 applied to instrumentation used for radiochemical analysis (e.g., gamma spectroscopy systems) if 1424 employing a method provided by the instrument manufacturer.

1425 1426 4.6 Calibration of Measuring and Test Equipment 1427 1428 Measuring and test equipment should be calibrated using NIST-traceable radioactive sources.

1429 The source geometries should be representative of the sample types analyzed and have the necessary 1430 accuracy, stability, and activity concentrations for their intended use.

1431 1432 4.7 Calibration Frequency 1433 1434 Calibrations should generally be performed at regular intervals in accordance with the frequencies 1435 established in NUREG-1301 and NUREG-1302. A change in calibration frequency (an increase or 1436 decrease) should be based on the reproducibility and time stability characteristics of the system. For 1437 example, an instrument system that gives a relatively wide range of readings when calibrated against a 1438 given standard should be recalibrated at more frequent intervals than one that gives measurements within 1439 a more-narrow range. Any monitoring system or individual measuring equipment should be recalibrated 1440 or replaced whenever it is suspected of being out of adjustment, excessively worn, or otherwise damaged 1441 and not operating properly.

1442 1443 4.8 Measurement Uncertainty 1444 1445 The measurement uncertainty (formerly called measurement error) associated with the 1446 measurement of radioactive materials in effluents should be estimated. Counting statistics can provide an 1447 estimate of the statistical counting uncertainty involved in radioactivity analyses. Because it may be 1448 difficult to assign error terms for each parameter affecting the final measurement, detailed statistical 1449 evaluations of error are not required. Normally, the statistical counting uncertainty decreases as the 1450 amount (concentration) of radioactivity increases. Thus, for the radioactive effluent release report, the 1451 statistical counting uncertainty is typically a small component of the total uncertainty. The sampling 1452 uncertainty is likely the largest component and includes uncertainties such as the uncertainty in 1453 volumetric and flow-rate measurements and laboratory processing uncertainties.

1454 1455 The total or expanded measurement uncertainty associated with the effluent measurement should 1456 ideally include the cumulative uncertainties resulting from the total operation of sampling and 1457 measurement. Expanded uncertainty should be reported with measurement results. The objective should 1458 be to evaluate only the important contributors and obtain a reasonable measure of the uncertainty 1459 associated with reported results. Detailed statistical and experimental evaluations are not required. The 1460 overall objective should be to obtain an overall estimate of measurement uncertainty. The formula for 1461 calculating the total or expanded uncertainty classically includes the square root of the sum of squares of 1462 each important contributor to the measurement uncertainty. Licensees may obtain additional information 1463 from NUREG-1576 and ANSI/HPS N13.1-2011.

1464 1465 4.9 Calibration of Accident-Range Radiation Monitors and Accident-Range Effluent Monitors 1466 1467 GDC 64 requires means for monitoring radioactivity in the reactor containment atmosphere; 1468 spaces containing components for recirculation of loss-of-coolant accident (LOCA) fluids; effluent 1469 discharge paths; and the plant environs for radioactivity that may be released from normal operations, 1470 including anticipated operational occurrences, and from postulated accidents. The regulation at 1471 10 CFR 20.1501(c) requires periodic calibration of instruments and equipment used to perform 1472 quantitative radiation measurements (e.g., dose rate and effluent monitoring).

1473

RG 1.21, Rev. 3, Page 33 1474 NUREG-0737, Item II.F.1, provides guidance for monitoring radiation levels and gaseous 1475 effluent during postulated radiological emergencies. RG 1.97, Revisions 2 and 3, provide guidance on the 1476 design and performance criteria of instrumentation used to assess plant and environ conditions during and 1477 following an accident. This RG 1.21 provides further guidance on the calibration of such instrumentation 1478 based on the NRCs Proposed Guidance for Calibration and Surveillance Requirements to Meet Item 1479 II.F.1 of NUREG-0737, issued August 1982 (Ref. 75). NUREG/CR-5569, Health Physics Positions 1480 Data Base, Health Physics Position (HPPOS)-001, Proposed Guidance for Calibration and Surveillance 1481 Requirements to Meet Item II.F.1 of NUREG-0737, issued February 1994 (Ref. 76), summarizes this 1482 additional guidance.

1483 1484 Noble Gas Monitoring - NUREG-0737, Item II.F.1-1, describes accident-range noble gas effluent 1485 monitors as monitors that are normally noble gas gross activity monitors sensitive to gamma emissions, 1486 beta emissions, or a mix of gamma and beta emissions. These monitors normally indicate (read out) in 1487 units of activity concentration, a count rate, or a dose rate (i.e., an indirect measurement of the noble gas 1488 gross activity concentration). Therefore, in order to determine the release rate of noble gas gross activity, 1489 a conversion factor (i.e., hereafter referred to as an instrument response factor) should be developed to 1490 convert the instrument output into an activity concentration for use in determining a release rate 1491 (e.g., curies per second of a mix of noble gases).

1492 1493 The initial vendor calibration of emergency effluent monitoring instruments may be a one-time 1494 prototype calibration based on the initial calibration of a single instrument of a certain model using 1495 NIST-traceable radiation sources. This initial prototype calibration of a single instrument model may 1496 include a determination of its fundamental characteristics, such as the following:

1497 1498

1.

a dose-rate linearity check using a radioactive gas or solid source (e.g., cesium (Cs)-137) to 1499 obtain three on-scale values separated by two decades of scale; 1500 1501

2.

a measurement of the instruments response factor to a calibration gas (e.g., xenon (Xe)-133 or 1502 krypton (Kr)-85);

1503 1504

3.

a characterization of the instruments energy -dependency characteristics, using solid sources 1505 ranging in gamma energy from low energy (e.g., 81 kiloelectron volts) to high energy 1506 (e.g., 2 megaelectron volts); and 1507 1508

4.

a determination, using a solid source, of a transfer factor that provides a dual purpose:

1509 1510

a.

for use by vendors to validate that subsequent instruments produced for sale of the same 1511 model have similar performance characteristics to the initial type instrument models 1512 characteristics; and 1513 1514

b.

for use by end users (e.g., nuclear power plants) in performing post installation and 1515 subsequent periodic calibration to verify that the instruments installed in the facility are 1516 functioning consistently with respect to initial vendor calibration of that instrument 1517 model.

1518 1519 Time-dependent (i.e., time since reactor shutdown) instrument response factors may be developed 1520 for each major accident type (i.e., a small-break LOCA with normal reactor coolant system activity levels, 1521 a large-break LOCA with gas gap activity levels, or a core-melt accident with noble gas activity levels 1522 arising from the fuel pellets release of noble gas). Each accident type has a characteristic, time-dependent 1523

RG 1.21, Rev. 3, Page 34 noble gas isotopic mix. In general terms, a small-break LOCA has a substantially decayed noble gas mix 1524 from the reactor coolant system with predominantly low-energy gamma photons; a large-break LOCA has 1525 a somewhat decayed noble gas mix from the gas gap of the fuel assemblies with predominately medium-1526 energy gamma photons; and a core -melt accident has a substantially undecayed mix of noble gas isotopes 1527 in the fuel pellets with predominately high-energy gamma photons.

1528 1529 The time-dependent instrument response factor accounts for the detectors energy efficiency at 1530 various gamma energies of the noble gas isotopic mix for that accident type. The instrument response 1531 factor normally has units of microcuries per cubic centimeter (µCi/cc) per count per minute or µCi/cc per 1532 milliroentgen per hour where the µCi/cc is the gross (total) summation of all the noble gas activities in the 1533 isotopic mix for each major type of accident listed above. It is also acceptable to use instrument response 1534 factors based on a single calibration gas with a low -energy gamma source (e.g., Xe-133) or beta 1535 emissions (e.g., Kr-85) for beta sensitive monitors.

1536 1537 The initial calibration process performed by the vendor does not need to be repeated at a nuclear 1538 power plant. Instead, a periodic single point source response check of the instruments performance as 1539 compared to a transfer factor provided by the vendor using a solid source - see ANSI N320-1978, 1540 Performance Specifications for Reactor Emergency Radiological Monitoring Instrumentation (Ref. 77).

1541 1542 Iodine and Particulate Monitoring - NUREG-0737, Item II.F.1-2, provides guidance on iodine 1543 and particulate effluent monitoring by sampling and analysis. Real-time monitoring is not required or 1544 considered practical; however, the licensees should have established procedures for collection of iodine 1545 and particulate samples and subsequent analysis to determine the release rate. For emergency dose 1546 assessment purposes, RASCAL (NUREG-1940 Section 1.2.8) can also be used to assess a real-time 1547 iodine and particulate release rate based on partitioning (scaling) factors to noble gases.

1548 1549 Containment High Range Monitoring - NUREG-0737, Item II.F.1-3, provides guidance on 1550 calibration of containment high-range monitors. An in-place calibration should be performed using a 1551 radioactive source at one point on the decade below 10 roentgens per hour (R/hr). Instrument scales in 1552 the range of 10 R/hr to 1E7 R/hr should be checked using electronic signal substitution with a calibrated 1553 current source to demonstrate that the system is functioning to higher radiation fields.

1554 1555 Containment high-range monitors should be used to assess the amount of core damage and to 1556 assess the source terms for the containment leakage release pathway. NUREG-1940, Section 1.2.4, 1557 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 1558 shutdown that correlates the containment radiation monitor readings to the amount of reactor damage for 1559 normal coolant, spiked coolant, cladding failure, and core melt accident scenarios.

1560 1561

5.

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

1564 The regulations referenced in Sections 5.4-5.6 of this RG contain both dose limits and design objectives 1565 that the licensee demonstrates compliance with through calculations. Table 1 summarizes the 1566 fundamental parameters associated with the dose calculations. RG Sections 5.7 and 5.8 present important 1567 concepts for these calculations. Because of differences between NRC and EPA regulations, 1568 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 35 demonstrating compliance only with radiological effluent technical specifications (based on 1569 10 CFR Part 50, Appendix I) does not necessarily ensure compliance with the EPAs 40 CFR Part 190, 1570 particularly if there is a direct radiation component (e.g., from BWR shine, ISFSI, or radioactive materials 1571 storage).

1572 1573 Table 1 - Parameters Associated with Dose Calculations 1574 1575 10 CFR PART 50, APPENDIX I 10 CFR 20.1301(e)

(EPA 40 CFR PART 190)

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

EPA 40 CFR Part 190 Where unrestricted area unrestricted area Individual Receptor real person/exposure pathway (nearest real residence, real garden, real dairy/meat animal) real person/exposure pathway (nearest real (residence, real garden, real dairy/meat 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 Material licensed only (per Appendix I,Section II radioactive materials - see Section 5.4 below) licensed and unlicensed (see Section 5.6 below)

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

1582 1583 For example, licensees may use conservative bounding dose assessments in lieu of site-specific 1584 dose assessments of the maximum dose to individual members of the public. Instead of assessing dose 1585 from ground-level effluent releases to a real individual member of the public located 3.2 km (2 miles) 1586 from the site boundary, a conservative bounding dose assessment can be performed for a hypothetical 1587 individual member of the public located at the site boundary.

1588 1589 If bounding assumptions are made, the radioactive effluent release report should state such and 1590 should annotate the assumptions. Hypothetical exposure pathways (see definition in the glossary) and 1591 locations are sometimes used for bounding dose assessments (or hazard evaluations done in accordance 1592 with 10 CFR 20.1501).

1593 1594

RG 1.21, Rev. 3, Page 36 5.2 Individual Members of the Public 1595 1596 Individual members of the public reside in the unrestricted area but at times may enter the 1597 controlled area of a commercial nuclear power plant. Each licensee is responsible for classifying 1598 individuals (by location) as either members of the public or as occupational workers (see the definition of 1599 members of the public in 10 CFR Part 20.) The NRC annual dose limits for members of the public are 1600 100 mrem total effective dose equivalent in accordance with 10 CFR 20.1301(a) and (b). In addition, in 1601 accordance with 10 CFR 20.1301(e) for uranium fuel cycle licensees (including nuclear power plants), the 1602 annual dose limits to members of the public are the EPA 40 CFR Part 190 limits of 25 mrem whole body, 1603 75 mrem to the thyroid, and 25 mrem to any other organ while in the unrestricted area.

1604 1605 If bounding assessments are not used, licensees should perform evaluations to determine the dose 1606 to a real, maximum exposed member of the public, regardless of whether the individual is in an 1607 unrestricted area or a controlled area. A member of the public is typically a real individual in a 1608 designated location where there is a real exposure pathway (e.g., a real garden, real cow, real goat, or 1609 actual drinking water supply) and not a fictitious fencepost resident or an exposure pathway that includes 1610 a virtual goat or cow. Licensees are encouraged (but not required) to use real individual members of the 1611 public when performing dose assessments for radioactive discharges. Table 1 in RG 1.109 allows a dose 1612 evaluation to be performed at a location where an exposure pathway and dose receptor actually existed at 1613 the time of licensing.

1614 1615 5.3 Occupancy Factors 1616 1617 For members of the public in the unrestricted area, occupancy factors should be assumed to be 1618 100 percent at locations identified in the land use census, unless site -specific information indicates 1619 otherwise. Occupancy factors may be applied inside the controlled area based on estimated hours spent in 1620 the controlled area.

1621 1622 5.4 10 CFR Part 50, Appendix I, Design Objectives and Limiting Conditions for Operation 1623 1624 Appendix I to 10 CFR Part 50 contains numerical guidance for design objectives and limiting 1625 conditions of operation for radioactive waste systems to ensure discharges of radioactive liquid and 1626 gaseous effluents to unrestricted areas are ALARA. This numerical guidance is listed in terms of annual 1627 air doses (gamma and beta), annual total body doses, and annual organ doses (see below). Licensee 1628 technical specifications require that exposure to liquid and gaseous effluents conform to the numerical 1629 guidance in 10 CFR Part 50, Appendix I. In accordance with 10 CFR 50.34a, these numerical guides for 1630 design objectives and limiting conditions of operation are not to be construed as radiation protection 1631 standards. For these dose calculations, the following terms are generally used:

1632 1633

1.

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

1634

2.

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

1635

3.

current annual period (excludes accumulated radioactivity from prior-year effluents), and 1636

4.

unrestricted area (excludes individuals in the restricted areas and controlled areas).

1637 1638 When calculating air doses, licensees should assure that, for any location outside the site 1639 boundary, doses do not exceed the dose objectives in 10 CFR Part 50, Appendix I. Calculation of air 1640 dose at the site boundary would assure the most conservative calculation of air doses for ground-level 1641 releases. This may not be true for elevated releases. Licensees should select a location that assures the 1642 most conservative calculation of air dose.

1643 1644

RG 1.21, Rev. 3, Page 37 5.5 10 CFR 20.1301(a) NRC dose limits for individual members of the public 1645 1646 This regulation specifies dose limits for members of the public from licensed operation of the 1647 facility. These limits apply to doses resulting from licensed and unlicensed radioactive material and from 1648 radiation sources other than background radiation (see 10 CFR 20.1001, Purpose). The dose limits 1649 include contributions to doses from (1) current-year effluents, (2) current-year direct radiation from the 1650 facility, and (3) accumulated radioactivity from prior-year effluents. The Technical Specifications 1651 establish the Radioactive Effluent Controls Program and the Environmental Monitoring Program, which 1652 establish effluent control methods sufficient to demonstrate of compliance with the NRC public dose 1653 limits in 10 CFR 20.1301(a).

1654 1655 5.6 10 CFR 20.1301(e) EPA Environmental Radiation Standards for the Uranium Fuel Cycle 1656 1657 For those facilities subject to the EPAs generally applicable environmental radiation standards in 1658 40 CFR Part 190, licensees must assess the highest cumulative (whole body and organ) doses from the 1659 uranium fuel cycle to a real individual in the general environment (i.e., outside the site boundary). The 1660 dose limits include contributions to doses from (1) current-year effluents, (2) current-year direct radiation 1661 from the facility, and (3) accumulated radioactivity from prior-year effluents. The Technical 1662 Specifications establish the Radioactive Effluent Controls Program and the Environmental Monitoring 1663 Program, which establish effluent control requirements sufficient to demonstrate compliance with the 1664 EPA public dose limits in 40 CFR Part 190 (see NUREG-0543).

1665 1666 These requirements include the following considerations:

1667 1668

1.

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

1669 1670

2.

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

1672 1673

3.

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

1676 1677

4.

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

1678 1679

5.

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

1681 1682

6.

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

1685 1686

7.

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

1688 1689

8.

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

1693 1694

RG 1.21, Rev. 3, Page 38 5.7 Dose Assessments for 10 CFR Part 50, Appendix I 1695 1696 Dose assessments to show compliance with technical specification requirements for meeting the 1697 numerical values of 10 CFR Part 50, Appendix I, design objectives should include quarterly and annual 1698 doses using the considerations in Section 5.4 of this RG. The dose assessments should be reported in a 1699 format similar to that shown in Table A-4 in Appendix A to this RG and include the items listed below:

1700 1701

1.

doses from liquid effluents 1702

a.

total body dose, quarterly and annually; 1703

b.

organ dose, quarterly and annually (maximum, any organ); and 1704

c.

percent of limits for each of the above.

1705 1706

2.

doses from gaseous effluents 1707

a.

beta and gamma air doses, quarterly and annually; 1708

b.

organ dose commitment from iodine, tritium, and particulate releases with half-lives 1709 greater than 8 days, quarterly and annually; and 1710

c.

percent of limit for each of the above.

1711 1712 An evaluation of the local exposure pathways to determine the maximum exposed member of the 1713 public should be performed. However, maximum doses from various exposure pathways are not additive 1714 from different locations. For example, dose from a downstream drinking water exposure pathway should 1715 not be added to the dose to an upstream resident whose exposure is from gaseous effluents and direct 1716 radiation unless that individuals drinking water is obtained from the downstream location.

1717 1718 Maximum doses to real individuals should be assessed as described in RG 1.109. The locations 1719 and exposure pathways are those where real individuals are present and exposed. Maximum exposed 1720 individuals are characterized as maximum with regard to food consumption, occupancy, and other 1721 usage in the vicinity of the plant site. For example, licensees should make maximum assumptions for 1722 food consumption and occupancy factors at actual locations when assessing dose to the maximum 1723 exposed individual, unless they have determined and applied site -specific (actual) data. In lieu of 1724 assessing dose to real individuals, licensee may also use bounding dose assessments for compliance with 1725 10 CFR Part 50, Appendix I (see the section titled Bounding Assessments).

1726 1727 The objective of 10 CFR Part 50, Appendix I, is to provide numerical guides for design objectives 1728 and limiting conditions for operation to ensure that radioactive effluent control equipment is effective in 1729 reducing emissions to ALARA levels. The numerical guidance pertains to quarterly and annual dose 1730 criteria at or beyond the unrestricted area from current-year effluent discharges. The calculations related 1731 to Appendix I do not include dose from radioactivity in prior-year, accumulated, effluent discharges 1732 (e.g., last years radioactivity remaining in lake water is excluded). However, the dose calculations for 1733 demonstrating compliance with the EPA limits do include accumulated radioactivity (see Section 5.8 of 1734 this RG).

1735 1736 For purposes of demonstrating compliance with dose criteria for limiting dose to a member of the 1737 public in unrestricted areas in accordance with Technical Specifications conforming to 10 CFR 50, 1738 Appendix I, the exposure pathways and routes of exposure identified in RG 1.109 should be considered.

1739 An evaluation of other exposure pathways (not included in dose assessments) should be performed and 1740 maintained for purposes of demonstrating compliance with the staff position on significant exposure 1741 pathways. Calculational procedures should be based on models and data such that the actual exposure of 1742 an individual through appropriate pathways is unlikely to be substantially underestimated. A new 1743 exposure pathway should be included in the demonstration of compliance if the calculated dose from that 1744 new exposure pathway exceeds 10 percent of the 10 CFR 50 Appendix I, Section II numerical guides on 1745

RG 1.21, Rev. 3, Page 39 design objectives. Bounding dose assessments as described in Section 5.1 of this RG may be used in 1746 evaluating the dose from any new significant exposure pathways.

1747 1748 Real exposure pathways are identified for routine discharges and direct radiation based on the 1749 results of the land use census. Dose calculations should typically be performed based on real exposure 1750 pathways. Conversely, dose assessments (i.e., surveillances and dose calculations) are not needed for 1751 exposure pathways that do not exist at a site. For example, if the land use census does not identify the 1752 existence of an ingestion exposure pathway involving a milk animal, the licensee is not required to assess 1753 that route of exposure for the ingestion exposure pathway. Similarly, if a licensee discharges liquid 1754 radioactive waste to a body of water (either surface water or groundwater) and that body of water is not 1755 used as a source of drinking water (either private or public), a drinking water assessment is not required.

1756 For purposes of reporting information in the ARERR, there is a distinction between dose assessments for 1757 10 CFR Part 50, Appendix I, and hazard assessments that may be conducted for onsite spills and leaks, as 1758 outlined in 10 CFR 20.1501 (where bounding estimates may be necessary). (See the discussion of 1759 bounding dose estimates in Section 5.1 of this RG.)

1760 1761 5.8 Dose Assessments for 10 CFR 20.1301(e) 1762 1763 To show compliance with 10 CFR 20.1301(e), dose assessments should be reported according to 1764 the generally applicable environmental radiation standards in 40 CFR Part 190, with consideration of 1765 Section 5.6 of this RG, and in a format similar to Table A-5 of Appendix A to this RG.

1766 1767 5.8.1 The following should be reported:

1768 1769 whole body dose to the maximum individual member of the public, 1770 thyroid dose to the maximum individual member of the public, 1771 dose to any other organ to the maximum individual member of the public, and 1772 percent of the applicable limit.

1773 1774 5.8.2 One means of demonstrating compliance with 40 CFR Part 190 is listed in Volume 42 of the 1775 Federal Register, page 2859, which states the following:

1776 1777 In the case of light water reactors, demonstrating conformance with 1778 Appendix I of 10 CFR 50 are generally adequate for demonstrating compliance 1779 with [EPA 40 CFR Part 190].

1780 1781 As a result, a licensee that (1) can demonstrate that external sources of direct radiation are 1782 indistinguishable from background and (2) demonstrates compliance with the numerical dose guidance of 1783 10 CFR Part 50, Appendix I, may cite the above reference as the basis for demonstrating compliance with 1784 40 CFR Part 190. The NRC provides additional guidance in NUREG-0543, Methods for Demonstrating 1785 Compliance with the EPA Uranium Fuel Cycle Standard (40 CFR Part 190).

1786 1787 However, licensees that (1) have external sources of direct radiation that are above background 1788 and (2) demonstrate compliance with the numerical dose guidance of 10 CFR Part 50, Appendix I, must 1789 also include sources of direct radiation from uranium fuel cycle operations (e.g., including direct radiation 1790 from the licensed facility and co-located or nearby nuclear power facilities, as appropriate).

1791 1792 5.8.3 The dose contributions from direct radiation may be estimated based on either (1) direct radiation 1793 measurements (e.g., thermoluminescent dosimeters, optically stimulated dosimeters, radiation 1794 detection instruments), (2) calculations, or (3) a combination of measurements and calculations.

1795

RG 1.21, Rev. 3, Page 40 When direct radiation dose is determined by measurement, RG 4.13 provides guidance on 1796 determining the dose to members of the public. Several sources contain additional information 1797 on background subtraction for environmental dosimeters (Refs. 29, 79, 80, and 81). Methods of 1798 determining dose from direct radiation to the maximum exposed individual member of the public 1799 may also include extrapolation methods.

1800 1801 Licensees must demonstrate compliance with 10 CFR 20.1301(e) for the generally applicable 1802 environmental radiation standards in 40 CFR Part 190. These include the concept of a total dose (to the 1803 whole body and to any organ) from all sources related to the uranium fuel cycle (such as adjacent or 1804 nearby nuclear power plants).

1805 1806 Contributions to the total dose from radioactive effluents (liquid and gaseous) and direct radiation 1807 should be included, if applicable. Other sources (e.g., accumulated radioactive materials in offsite ponds 1808 or lakes from previous years discharges) should also be included, if applicable, when estimating the total 1809 dose. However, if the contributions from direct radiation or accumulated radioactivity are generally 1810 minor (as evaluated and documented in a licensee technical evaluation as not contributing to the total 1811 dose), these contributions need not be included in the total dose evaluation, but the basis for exclusion 1812 should be documented.

1813 1814 5.9 Dose Calculations 1815 1816 Acceptable dose assessment models, such as those provided in RGs 1.109, 1.111, 1.112, and 1817 1.113, should be used to make dose calculations. When calculating organ doses from airborne effluents 1818 for purposes of demonstrating compliance with Technical Specifications conforming to 10 CFR 50, 1819 Appendix I, contributions from I-131, I-133, tritium, and radionuclides in particulate form with half-lives 1820 greater than 8 days should be included in the assessment. For demonstrating compliance with EPA 40 1821 CFR 190, in addition to the above radionuclides, doses from C-14 should be included in organ dose 1822 assessments.

1823 1824

6.

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

1829 1830 Solid radioactive waste shipments should be reported in a format similar to that of Table A-3 in 1831 Appendix A to this RG. The total curie quantity and major radionuclides in the solid waste shipped off 1832 site should be determined and reported.

1833 1834 The data should be divided by the waste stream categories listed in Table A-3. The waste streams 1835 are:

1836 (1) wet radioactive waste (e.g., spent resin, filters, sludges, etc.),

1837 (2) dry radioactive waste (e.g., trash, paper, discarded protective clothing etc.),

1838 (3) activated or contaminated radioactive material (e.g., equipment or bulk radioactive material, 1839 etc.), and 1840 (4) other radioactive waste (waste not included in the above categories and not excepted from 1841 reporting as described below).

1842 1843 Shipments that do not need to be reported include shipments of metal melt, contaminated 1844 equipment for transfer between licensees or equipment for refurbishment, contaminated laundry (either 1845 launderable or dissolvable), or radioactive samples for analysis. Potentially contaminated dry active 1846

RG 1.21, Rev. 3, Page 41 waste sent for resurvey and segregation (sometimes referred to as green is clean) does not need to be 1847 reported. Equipment shipped for decontamination and free release does not need to be reported.

1848 However, records of these types of shipments should be maintained on site.

1849 1850 Note 1: Data on LLW disposed in licensed LLW disposal facilities is available using the Manifest 1851 Information Management System operated by the U.S. Department of Energy.

1852 1853 Note 2: There are no requirements for reporting storage of LLW at nuclear power plants.

1854 However, LLW storage records should be established and maintained at nuclear plants and made 1855 available for NRC inspection during routine effluent inspections consistent with applicable NRC 1856 requirements.

1857 1858 1859

7.

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

1862 1863 7.1 Examples of Small Errors 1864 1865 Small errors may be any of the following:

1866 1867

1.

inaccurate reporting of dose that equates to 10 percent of the applicable 10 CFR Part 50, 1868 Appendix I, design objective or 10 percent of the EPA public dose criterion; 1869 1870

2.

inaccurate reporting of curies (or release rates, volumes, etc.) that equate to 10 percent of the 1871 affected curie total (or release rate, volume, etc.) after correction; 1872 1873

3.

omissions that do not impede the NRCs ability to adequately assess the information supplied by 1874 the licensee; or 1875 1876

4.

typographical errors or other errors that do not alter the intent of the report.

1877 1878 7.2 Reporting Small Errors 1879 1880 Licensees should correct small errors within 1 year of discovery and may submit the correction 1881 with the next (normally scheduled) submittal of the ARERR, as follows. A brief narrative explanation of 1882 the errors should be included in Section 8, Errata/Corrections to Previous ARERRs, of Table A-6. The 1883 narrative should state that the affected pages, in their entirety, are included as attachments to the ARERR.

1884 Additionally, the corrected pages, in their entirety, should be submitted as an attachment (or addendum) 1885 to the ARERR. The corrected pages should reference the affected calendar year and should contain 1886 revision bars in the margins of the page to indicate the locations of the changes. If submitting corrections 1887 to multiple ARERRs, a separate attachment (or addendum) should be made for each of the affected years.

1888 Other methods of correcting previous ARERRs may be used, provided the corrections are clearly and 1889 completely described.

1890 1891 7.3 Examples of Large Errors 1892 1893 Large errors may be any of the following:

1894 1895

1.

inaccurate reporting of dose that equates to >10 percent of the 10 CFR Part 50, Appendix I, or 1896 EPA public dose criterion, after correction; 1897

RG 1.21, Rev. 3, Page 42 1898

2.

inaccurate reporting of curies (or release rate, volume, etc.) that equates to >10 percent of the 1899 affected curie total (or release rate, volume, etc.) after correction; 1900 1901

3.

omissions that may impede the NRCs ability to adequately assess the information supplied by 1902 the licensee; or 1903 1904

4.

typographical errors or other errors that significantly alter the intent of the report.

1905 1906 7.4 Reporting Large Errors 1907 1908 Licensee should correct large errors within 90 days of discovery. The correction may be made by 1909 special submittal or may be submitted with the next (normally scheduled) ARERR (if the next ARERR is 1910 to be submitted within 90 days of discovery of the error). If corrections are made by special submittal, 1911 the licensee should include a brief narrative explaining the errors. The narrative should state that the 1912 affected pages, in their entirety, are included as an attachment. The corrected pages should be attached in 1913 their entirety. The corrected pages should reference the affected calendar year and should contain 1914 revision bars in the margins of the page to indicate the locations of the changes. If submitting corrections 1915 to multiple ARERRs, separate attachment (or addendum) should be made for each of the affected years.

1916 If corrections are made coincident with the next (normally scheduled) submittal of the ARERR, the 1917 correction process should be used as specified in Section 7.2 (for small errors). Other methods of 1918 correcting previous ARERRs may be used provided the corrections are clearly and completely described 1919 consistent with NRC requirements on the completeness and accuracy of information.

1920 1921

RG 1.21, Rev. 3, Page 43 1922

8.

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

1926 The Technical Specifications establish a specific review and approval process for making changes to the 1927 ODCM. Potential changes require licensee analyses or evaluations justifying the change and a 1928 determination that the changes maintain the levels of radioactive effluent control required 1929 by10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I. The evaluation of potential 1930 changes should also consider the need for monitoring in support of decommissioning planning during 1931 operations (see RG 4.22).

1932 1933 Effluent and environmental monitoring programs may need to be modified once power operations 1934 have permanently ceased and a written certification has been submitted to the NRC in accordance with 1935 10 CFR 50.82, Termination of License. The evaluation of potential changes should consider the need 1936 for effluent and environmental monitoring during active decommissioning which is likely to affect 1937 principal release points and principal radionuclides. For example, the removal of effluent ventilation 1938 systems will likely change principal release points and there may be new principal radionuclides 1939 identified (e.g., Kr-85), while radioactive decay may have eliminated former principal radionuclides (e.g.,

1940 I-131) (see Section C.1). Potential changes must be reviewed and approved by the plant manager, station 1941 manager, or as described in plant-specific Technical Specifications, with submittal to the NRC as part of 1942 the next Annual Radioactive Effluent Release Report.

1943 1944 If the plant has a 10 CFR Part 72 ISFSI, the licensee must maintain compliance with the requirements 1945 in 10 CFR Part 72 regarding controls of effluent(s) and an environmental monitoring program. These 1946 requirements include 10 CFR 72.44(d) for 10 CFR Part 72 specific license ISFSIs and, for 1947 10 CFR Part 72 general license ISFSIs, any requirements specified in technical specifications of the 1948 certificate(s) of compliance for the storage systems in use at the ISFSI (to comply with 1949 10 CFR 72.212(b)(3) and (b)(5)).

1950 1951 The radiological criteria for license termination are addressed in 10 CFR 20 Subpart E. The 1952 radiological criteria for unrestricted use (10 CFR 20.1402) encompass contributions from residual 1953 radioactivity in soils and remnant site components and in groundwater. While some reductions in 1954 monitoring programs may be possible when operations cease, other aspects of monitoring such as 1955 groundwater monitoring may need to be increased to adequately characterize residual radioactivity and 1956 characterize dispersion pathways to support dose assessments and to estimate the decommissioning costs.

1957 Lessons learned documented in RG 1.185 and NUREG-1757 indicate that the monitoring data from the 1958 period of operation tend to be insufficient to allow the staff to fully understand the types and the 1959 movement of radioactive material contamination in groundwater at the facility, as well as the extent of the 1960 residual radioactivity. Decommissioning reporting and recordkeeping requirements are addressed in 1961 10 CFR 50.75(g).

1962 1963 Further general guidance to facilitate planning for decommissioning of power plants and 1964 facilities during operations can be found in RG 4.22, in RG 1.185 for post-shutdown decommissioning 1965 activities, in NUREG-1757 for consolidated decommissioning guidance, and in NUREG-1575, Rev. 1, 1966 Multi-Agency Radiation Survey and Site Investigation Manual.

1967 1968 1969 1970

9. Format and Content of the Annual Radioactive Effluent Release Report 1971 1972

RG 1.21, Rev. 3, Page 44 In accordance with 10 CFR 50.4, Written communications, licensees should submit their 1973 annual report electronically or in a written communication. The report should consist of a summary of the 1974 numerical data in a tabular format similar to Tables A-1-A-5 in Appendix A to this RG. Effluent data 1975 reported in Tables A-1, A-1A - A-1F, A-2, A-2A, A-2B, and A-4 should be summarized on a quarterly 1976 and annual basis. Tables A-3 and A-5 should be summarized on an annual basis. In addition to 1977 numerical data, the report should include additional supplemental information containing all the 1978 information in (but not necessarily in the format of Table A-6). Additional detail for the information 1979 contained in each of these tables is listed below. To comply with 10 CFR 50.36a, licensee must submit 1980 their ARERR by May 1 (unless a licensing basis exists for a different submittal date) to report on 1981 effluents and solid waste from the previous calendar year.

1982 1983 Radionuclides that are not detected do not need to be listed in the tables (Tables A-1A-A-1F, A-1984 2A, and A-2B). Activity that is detected should be reported in the appropriate tables (i.e., Tables A-1, A-1985 2, A-1A - A-1F, A-2A, and A-2B) in the ARERR, provided that the amount discharged is numerically 1986 significant with respect to the three-digit exponential format recommended for the ARERR. This should 1987 not be confused with three significant figures. Licensees may round numbers according to accepted 1988 practices (e.g., refer to ASTM E-29, Standard Practice for Using Significant Digits in Test Data to 1989 Determine Conformance with Specifications (Ref. 87)); however, after rounding has been completed, 1990 values should be reported in the ARERR in a three-digit exponential format. Measurements should be 1991 reported for positive values. Some radionuclides that are detected in a year may not be detected in all 1992 quarters. If results are determined to be below detectable levels for an entire quarter, the table entry 1993 should include a suitable designation (e.g., N/D (not detected) and an accompanying footnote) to denote 1994 that measurements were performed but activity was not detected.

1995 1996 The format specified in this RG revision differs slightly from the format specified in Revision 1 1997 and Revision 2. The format and content specified in this Revision 3 of RG 1.21 is one acceptable method 1998 of reporting the data. Other formats may be used (e.g., some tables may be combined) as long as the 1999 specified content is provided (e.g., quarterly totals and annual totals by each release category). However, 2000 licensees are encouraged to use the format listed below to maximize consistency in data reporting. This 2001 format is designed to be consistent with some commonly used electronic-data-reporting software 2002 packages. Consistency in reporting format aids review by members of the public and allows easier 2003 industrywide comparisons of the data.

2004 2005 10 CFR 72 licensees may also, if they choose to do so, use the format specified in this RG for 2006 independent spent fuel storage installation (ISFSI) effluent reports required by 10 CFR 72.44(d) (for 2007 specific licenses) or the storage system(s) certificate(s) of compliance (for general licenses). However, 2008 the ISFSI effluent reporting requirement is not normally satisfied by inclusion as part of the ARERR 2009 since the reporting dates may conflict. If the dates are coincident, or can be met with a single report, 2010 licensees may use the ARERR to fulfill the ISFSI reporting requirements, provided the licensee submits a 2011 copy as specified in those requirements (e.g., 10 CFR 72.44(d)(3) for specific licenses).

2012 2013 9.1 Gaseous Effluents 2014 2015 The quarterly and annual sums of all radionuclides discharged in gaseous effluents (i.e., routine 2016 and abnormal discharges, continuous, and batch) should be reported in a format similar to that of 2017 Tables A-1A - A-1F in Appendix A to this RG. The data should then be further summarized and 2018 reported in the format of Table A-1.

2019 2020 Table A-1, Gaseous EffluentsSummation of All Discharges, contains a summation of all 2021 gaseous effluent discharges from all release points and all modes of release. The data are subdivided by 2022

RG 1.21, Rev. 3, Page 45 quarter and year for each radionuclide category: fission and activation gases, iodines/halogens, 2023 particulates, tritium, gross alpha and carbon-14.

2024 Table A-1A, Gaseous EffluentsGround-Level ReleaseBatch Mode, contains a summation 2025 of gaseous effluent releases from ground-level release points in the batch mode of release for six 2026 radionuclide categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha 2027 and carbon-14. Licensees should report the following:

2028 2029

1.

curies of each radionuclide discharged by quarter and year, and 2030 2031

2.

total curies discharged in each radionuclide category by quarter and year.

2032 2033 Some licensees may have surveillance requirements allowing the non-noble gas radionuclides 2034 (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with 2035 continuous release results. In these instances, the table entries for the affected radionuclides for batch 2036 releases should include an appropriate designation (e.g., *) and an accompanying footnote describing 2037 this situation.

2038 2039 Table A-1B, Gaseous EffluentsGround-Level ReleaseContinuous Mode, contains a 2040 summation of gaseous effluent releases from ground-level release points in the continuous mode of 2041 release for six radionuclide categories: fission and activation gases, iodines/halogens, particulates, 2042 tritium, gross alpha and carbon-14. Licensees should report the following:

2043 2044

1.

curies of each radionuclide discharged by quarter and year, and 2045

2.

total curies discharged in each radionuclide category by quarter and year.

2046 2047 Table A-1C, Gaseous EffluentsElevated ReleaseBatch Mode, contains a summation of 2048 gaseous effluent releases from elevated release points in the batch mode of release for six radionuclide 2049 categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha, and carbon-2050

14. Licensees should report the following:

2051 2052

1.

curies of each radionuclide released by quarter and year, and 2053

2.

total curies released in each radionuclide category by quarter and year.

2054 2055 Some licensees may have surveillance requirements allowing the non-noble gas radionuclides 2056 (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with 2057 continuous release results. In these instances, the table entries for the affected radionuclides for batch 2058 releases should include an appropriate designation (e.g., *) and an accompanying footnote describing 2059 this situation.

2060 2061 Table A-1D, Gaseous EffluentsElevated ReleaseContinuous Mode, contains a summation 2062 of gaseous effluent releases from elevated release points in the continuous mode of release for six 2063 radionuclide categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha 2064 and carbon-14. Licensees should report the following:

2065 2066

1.

curies of each radionuclide released by quarter and year, and 2067 2068

2.

total curies released in each radionuclide category by quarter and year.

2069 2070 Table A-1E, Gaseous EffluentsMixed Mode ReleaseBatch Mode, contains a summation of 2071 gaseous effluent releases from mixed-mode release points in the continuous mode of release for six 2072

RG 1.21, Rev. 3, Page 46 radionuclide categories: fission and activation gases, iodines/halogens, particulates, tritium, gross alpha, 2073 and carbon-14. Licensees should report the following:

2074 2075

1.

curies of each radionuclide released by quarter and year, and 2076 2077

2.

total curies released in each radionuclide category by quarter and year.

2078 2079 Some licensees may have surveillance requirements allowing the non-noble gas radionuclides 2080 (e.g., iodines and tritium) for some types of batch releases (e.g., containment purge) to be reported with 2081 continuous release results. In these instances, the table entries for the affected radionuclides for batch 2082 releases should include an appropriate designation (e.g., *) and an accompanying footnote describing 2083 this situation.

2084 2085 Table A-1F, Gaseous EffluentsMixed Mode ReleaseContinuous Mode, contains a 2086 summation of gaseous effluent releases from mixed-modes release points in the continuous mode of 2087 release for six radionuclide categories: fission and activation gases, iodines/halogens, particulates, 2088 tritium, gross alpha, and carbon-14. Licensees should report the following:

2089 2090

1.

curies of each radionuclide released by quarter and year, and 2091 2092

2.

total curies released in each radionuclide category by quarter and year.

2093 2094 9.2 Liquid Effluents 2095 2096 The quarterly and annual sums of all radionuclides discharged in liquid effluents (i.e., routine and 2097 abnormal discharges, continuous, and batch) should be reported in a format similar to that of Tables A-2A 2098 and A-2B. The data should then be further summarized and reported in the format of Appendix A, 2099 Table A-2.

2100 2101 Table A-2, Liquid EffluentsSummation of All Releases, contains a summation of all liquid 2102 radioactive discharges from all release points and all modes of release. The data are subdivided by 2103 quarter and year for each of the radionuclide categories: fission and activation products, tritium, 2104 dissolved and entrained noble gases, and gross alpha.

2105 2106 The table also includes the total volume of primary coolant waste (typically batch mode 2107 releases) before dilution. In this context, primary coolant waste means the higher activity waste that 2108 generally is not discharged directly but is instead typically processed through the liquid radioactive waste 2109 treatment system before discharge. Various methods exist for calculating the dilution water flow rate.

2110 HPPOS-099, Attention to Liquid Dilution Volumes in Semiannual Radioactive Effluent Release 2111 Reports, issued November 1984 (Ref. 88), indicates that licensees should use the total volume of dilution 2112 flow, not just that flow during periods of liquid effluent releases. Licensees should include information 2113 describing how this value is calculated in either the ODCM or the ARERR. Because the primary coolant 2114 waste typically accounts for the vast majority of the radioactivity in liquid waste discharges, the NRC 2115 recommends that the volume and dilution data be summarized separately from the low-activity waste 2116 described in the following paragraph.

2117 2118 The total measured volume or average flow rate of waste from secondary or balance-of-plant 2119 systems (e.g., steam generator blowdown, low-activity waste sumps, and auxiliary boilers) should be 2120 reported. In this context, secondary or balance-of-plant waste means the typically very low-activity waste 2121 that is generally not processed with the liquid radioactive waste treatment system and that collectively 2122 represents a very large volume of waste. Various methods exist for calculating the dilution water flow 2123

RG 1.21, Rev. 3, Page 47 rate. HPPOS-099 states that licensees should use the total volume of dilution flow, not just that volume 2124 discharged during periods of liquid effluent releases. Licensees should include information describing 2125 how this value is calculated in either the ODCM or the ARERR. Because of the potentially high volume 2126 and extremely low activity of this type of waste, the NRC recommends the volume and dilution data be 2127 summarized separately from the higher activity waste described in the previous paragraph.

2128 2129 Licensees should report dilution flow rates during periods of release (before effluent is discharged 2130 to the receiving water body), as described above. If calculated differently than described above, the 2131 licensee should describe the method of calculation. Licensees may choose to report near-field dilution if 2132 they account for dilution by the receiving water body. Licensees may report the average, minimum, peak 2133 river, and stream flow rates, as applicable.

2134 2135 Table A-2A, Liquid EffluentsBatch Mode, contains a summation of liquid effluent 2136 discharges in the batch mode of release. The table is divided into four radionuclide categories: fission 2137 and activation products, tritium, dissolved and entrained gases, and gross alpha. Licensees should report 2138 the following:

2139 2140

1.

curies of each radionuclide and gross alpha discharged by quarter and year, and 2141 2142

2.

total curies in each radionuclide category by quarter and year.

2143 2144 Table A-2B, Liquid EffluentsContinuous Mode, contains a summation of liquid effluent 2145 discharges in the continuous mode of release. The table is divided into four radionuclide categories:

2146 fission and activation products, tritium, dissolved and entrained gases, and gross alpha. Licensees should 2147 report the following:

2148 2149

1.

curies of each radionuclide and gross alpha discharged by quarter and year, and 2150 2151

2.

total curies in each radionuclide category by quarter and year.

2152 2153 9.3 Solid Waste Shipments Released from the Unit (per Standard Technical Specifications) 2154 2155 Appendix A, Table A-3, provides an acceptable format for reporting the solid radioactive waste 2156 released (shipped) from the unit (plant site) during the reporting period. The NRC intends that licensees 2157 report the waste shipped from the site, regardless of whether the shipment is sent for waste processing or 2158 direct disposal (i.e., with or without waste processing).

2159 2160 Licensee should report the volume and curies of solid waste shipped (see exceptions noted in 2161 Section 6) for each of the following waste streams:

2162 2163

1.

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

2164 2165

2.

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

2166 2167

3.

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

2168 and 2169 2170

4.

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

2172 2173

RG 1.21, Rev. 3, Page 48 9.4 Dose Assessments 2174 2175 Licensees should calculate the annual evaluations of dose to members of the public using 2176 RG 1.21, Section 5 and report the data in the format of Tables A-4 and A-5. Dose assessments should 2177 demonstrate compliance with the following9:

2178 2179

1.

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

2181 2182

a.

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

2187 2188

b.

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

2191 2192

c.

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

2195 2196

2.

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

2198 2199

a.

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

2202 2203 9.5 Supplemental Information 2204 2205 Licensees should provide supplemental information in a descriptive, narrative form (see 2206 Table A-6 or in a similar format). Relevant information and a description of circumstances should be 2207 provided as appropriate for each the following categories, adding categories as appropriate. The 2208 annotation N/A should be used if a category is not applicable.

2209 2210 9.5.1 Abnormal Releases or Abnormal Discharges 2211 2212 The reporting of abnormal releases to onsite areas and abnormal discharges to unrestricted areas 2213 should include the following:

2214 2215

1.

Specific information should be reported concerning abnormal (airborne, liquid) releases on site 2216 and abnormal discharges to the unrestricted area. The report should describe each event in a way 2217 that would enable the NRC to adequately understand how the material was released and if there 2218 was a discharge to the unrestricted area. The report should describe the potential impact on the 2219 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 49 ingestion exposure pathway involving surface water and groundwater, as applicable. The report 2220 should also describe the impact (if any) on other affected exposure pathways (e.g., inhalation).

2221 2222

2.

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

2224 2225

a.

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

2228 2229

b.

abnormal releases or abnormal discharges estimated to exceed 300 liters (100 gallons) of 2230 radioactive liquid where the presence of licensed radioactive material is positively 2231 identified (in either the onsite environs or in the source of the leak or spill) as greater than 2232 the minimum detectable activity11 for the laboratory instrumentation; 2233 2234

c.

abnormal releases to onsite areas that result in detectable residual radioactivity after 2235 remediation; 2236 2237

d.

abnormal releases that result in a high effluent radiation alarm without an anticipated 2238 system trip occurring; and 2239 2240

e.

abnormal discharges to an unrestricted area.

2241 2242

3.

Information on abnormal releases or abnormal discharges should include the following, as 2243 applicable:

2244 2245

a.

date and duration, 2246

b.

location, 2247

c.

volume, 2248

d.

estimated activity of each radionuclide, 2249

e.

effluent monitoring results (if any),

2250

f.

onsite monitoring results (if any),

2251

g.

depth to the local water table, 2252

h.

classification(s) of subsurface aquifer(s) (e.g., drinking water, unfit for drinking water, 2253 not used for drinking water),

2254

i.

size and extent of any groundwater plume, 2255

j.

expected movement/mobility of any groundwater plume, 2256

k.

land use characteristics (e.g., water used for irrigation),

2257

l.

remedial actions considered or taken and results obtained, 2258

m.

calculated member of the public dose attributable to the release, 2259

n.

calculated member of the public dose attributable to the discharge, 2260

o.

actions taken to prevent recurrence, as applicable, and 2261

p.

whether the NRC was notified, the date(s), and the contact organization.

2262 2263 9.5.2 Nonroutine Planned Discharges 2264 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 50 Discharges resulting from remediation efforts that are not identified in the ODCM should be 2266 reported. For example, the remediation effort may include pumping of contaminated groundwater in 2267 response to leaks and spills.

2268 2269 9.5.3 Radioactive Waste Treatment System Changes 2270 2271 Licensees should report any changes or modifications affecting any portion of the gaseous 2272 radioactive waste treatment system, the ventilation exhaust treatment system, or the liquid radioactive 2273 waste treatment.

2274 2275 9.5.4 Annual Land Use Census Changes 2276 2277 Licensees should report any changes or modifications affecting significant aspects of the 2278 environmental monitoring program such as receptors, receptor locations, sample media availability, or 2279 new (or changed) routes of exposure.

2280 2281 9.5.5 Effluent Monitoring System Inoperability 2282 2283 Licensees should report information on inoperable effluent monitors as follow:

2284 2285

1.

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

2291 2292

2.

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

2296 2297 9.5.6 Offsite Dose Calculation Manual Changes 2298 2299 Licensees should report any changes or modifications affecting significant aspects of the ODCM.

2300 2301 9.5.7 Process Control Program Changes 2302 2303 Licensees should report any changes or modifications affecting significant aspects of the ODCM.

2304 2305 9.5.8 Corrections to Previous Reports 2306 2307 When submitting corrections to previous reports, licensees should do the following:

2308 2309

1.

Include a brief explanation of the error(s).

2310 2311

2.

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

2312 2313

3.

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

2316

RG 1.21, Rev. 3, Page 51 2317 9.5.9 Other (Narrative Descriptions of Other Information Related to Radioactive Effluents) 2318 2319 Licensees should report other supplemental information (as appropriate).

2320

RG 1.21, Rev. 3, Page 52 D. IMPLEMENTATION 2321 2322 The NRC staff may use this regulatory guide as a reference in its regulatory processes, such as 2323 licensing, inspection, or enforcement. However, the NRC staff does not intend to use the guidance in this 2324 regulatory guide to support NRC staff actions in a manner that would constitute backfitting as that term is 2325 defined in 10 CFR 50.109, Backfitting, or in 10 CFR 72.62, Backfitting, and as described in NRC 2326 Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information 2327 Requests, (Ref. 90), nor does the NRC staff intend to use the guidance to affect the issue finality of an 2328 approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The 2329 staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes 2330 forward fitting as that term is defined and described in Management Directive 8.4. If a licensee believes 2331 that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this 2332 Implementation section, then the licensee may file a backfitting or forward fitting appeal with the NRC in 2333 accordance with the process in Management Directive 8.4.

2334 2335 2336 2337

RG 1.21, Rev. 3, Page 53 GLOSSARY 2338 2339 a prioriBefore-the-fact limit, representing the capability of a measurement system and not as an 2340 after-the-fact (a posteriori) limit for a particular measurement.

2341 2342 abnormal dischargeThe unplanned or uncontrolled emission of an effluent (i.e., containing 2343 plant-related, licensed radioactive material) into the unrestricted area.

2344 2345 abnormal releaseThe unplanned or uncontrolled emission of an effluent (i.e., containing plant-related, 2346 licensed radioactive material) into the onsite environs.

2347 2348 accumulated radioactivityRadioactivity from prior-year effluent releases that may still be present in 2349 the media of concern.

2350 2351 background (radiation)Means radiation from cosmic sources; naturally occurring radioactive 2352 material, including radon (except as a decay product of source or special nuclear material); and 2353 global fallout as it exists in the environment from the testing of nuclear explosive devices and 2354 from past nuclear accidents, such as Chernobyl, that contribute to background radiation and are 2355 not under the control of the licensee. Background radiation does not include radiation from 2356 source, byproduct, or special nuclear materials regulated by the Commission.

2357 2358 batch releaseThe release of liquid (radioactive) wastes of a discrete volume or the release of a tank or 2359 purge of radioactive gases into the site environs.

2360 2361 channel checkThe qualitative assessment of channel behavior during operation by observation. This 2362 determination should include, where possible, comparison of the channel indication, status with 2363 other indications, and status derived from independent instrument channels measuring the same 2364 parameter.

2365 2366 channel operational testThe injection of a simulated signal into the channel as close to the sensor as 2367 practicable to verify operability of alarm, interlock, and trip functions, as applicable. The channel 2368 operational test should include adjustments, as necessary, of the alarm, interlock, and trip 2369 setpoints, as applicable, such that the setpoints are within the required range and accuracy.

2370 2371 continuous releaseAn essentially uninterrupted release of gaseous or liquid effluent for extended 2372 periods during normal operation of the facility where the volume of radioactive waste is 2373 non-discrete and there is input flow during the release.

2374 2375 controlled area (10 CFR Part 20)An area outside of a restricted area but inside the site boundary, 2376 access to which is limited by the licensee for any reason.

2377 2378 controlled area (10 CFR Part 72)The area immediately surrounding an ISFSI or a monitored 2379 retrievable storage installation (MRS) for which the licensee exercises authority over its use and 2380 within which ISFSI or MRS operations are performed.

2381 2382 controlled dischargeA radioactive discharge is considered to be controlled if (1) the discharge was 2383 conducted in accordance with methods, and without exceeding any of the limits, outlined in the 2384 ODCM or (2) if one or more of the following three items are true:

2385 2386

RG 1.21, Rev. 3, Page 54

1.

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

2389 2390

2.

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

2393 2394

3.

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

2398 2399 controlled releaseA radioactive release is considered to be controlled if (1) the release was 2400 conducted in accordance with methods, and without exceeding any of the limits, outlined in the 2401 ODCM or (2) if one or more of the following three items are true:

2402 2403

1.

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

2406 2407

2.

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

2410 2411

3.

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

2415 2416 conversion factorA factor (e.g., microcuries per cubic centimeter per counts per minute (Ci/cc/cpm) 2417 used to estimate a radioactivity concentration in an effluent based on a gross radioactivity 2418 measurement (e.g., cpm).

2419 2420 D/QA deposition value used for estimating the dose to an individual at a specified (e.g., controlling) 2421 location. D/Q may be described as the downwind surface or ground deposition (D) (e.g., in units 2422 of microcuries per square meter [Ci/m2]) of radioactive material at a location, divided by the 2423 release activity (Q) (e.g., in Ci). D/Q is thus a normalized downwind surface deposition value 2424 per unit release and can be used to determine the surface or ground radioactivity concentration 2425 during a measured effluent release over a specific period of time. The units of D/Q are reciprocal 2426 square meters.

2427 2428 determinationA quantitative evaluation of the release or presence of radioactive material under a 2429 specific set of conditions. A determination may be made by direct measurement or indirect 2430 measurements (e.g., with the use of scaling factors).

2431 2432 dilution water (for liquid radioactive waste)For purposes of this RG, any water other than the 2433 undiluted radioactive waste that is mixed with undiluted liquid radioactive waste before its 2434 ultimate discharge to the unrestricted area.

2435 2436

RG 1.21, Rev. 3, Page 55 discharge pointA location at which radioactive material enters the unrestricted area. This would be 2437 the point beyond the vertical plane of the unrestricted area (surface or subsurface).

2438 2439 drinking waterWater that does not contain an objectionable pollutant, contamination, minerals, or 2440 infective agent and is considered satisfactory for domestic consumption. This is sometimes called 2441 potable water. Potable water is water that is safe and satisfactory for drinking and cooking.

2442 Although EPA regulations only apply to public drinking water sources supplying 25 or more 2443 people (refer to the EPA for more information), for purposes of the effluent and environmental 2444 monitoring programs, the term drinking water includes water from single-use residential drinking 2445 water wells.

2446 2447 effluentLiquid or gaseous waste containing plant-related, licensed radioactive material, emitted at the 2448 boundary of the facility (e.g., buildings, end-of-pipe, stack, or container) as described in the final 2449 safety analysis report.

2450 2451 effluent dischargeThe portion of an effluent release that reaches an unrestricted area. (See also the 2452 definition for radioactive discharge.)

2453 2454 effluent releaseThe emission of an effluent. (See also the definition for radioactive release.)

2455 2456 elevated releaseA gaseous effluent release made from a height that is more than twice the height of 2457 adjacent solid structures, or releases made from heights sufficiently above adjacent solid 2458 structures such that building wake effects are minimal or absent.

2459 2460 exposure pathwayA mechanism by which radioactive material is transferred from the (local) 2461 environment to humans. There are three commonly recognized exposure pathways: inhalation, 2462 ingestion, and direct radiation. For example, ingestion may include dose contributions from one 2463 or more routes of exposure. One route of exposure that may contribute to the ingestion exposure 2464 pathway is often referred to as grass-cow-milk-infant-thyroid route of exposure.

2465 2466 general environmentAn EPA 40 CFR 190.02 definition meaning the total terrestrial, atmospheric and 2467 aquatic environment outside sites upon which any (licensed) operation of a nuclear fuel cycle is 2468 conducted.

2469 2470 ground-level releaseA gaseous effluent release made from a height that is ator less thanthe height 2471 of adjacent solid structures, or where the degree of plume rise is unknown or is otherwise 2472 insufficient to avoid building wake effects.

2473 2474 groundwaterAll water in the surface soil, the subsurface soil, or any other subsurface water.

2475 Groundwater is simply water in the ground regardless of its quality, including saline, brackish, or 2476 fresh water. Groundwater can be moisture in the ground that is above the regional water table in 2477 the unsaturated (or vadose) zone, or groundwater can be at and below the water table in the 2478 saturated zone.

2479 2480 hypothetical exposure pathwayAn exposure pathway in which one or more of the components 2481 involved in the transfer of a radionuclide from the environment to the human does not actually 2482 exist at the specified location, or if a real human does not consume, inhale, or otherwise become 2483 exposed to the radioactive material. For example, the grass-cow-milk-infant-thyroid route of 2484 exposure (associated with the ingestion exposure pathway) would be considered a hypothetical 2485 exposure pathway if the grass, the cow, or the milk did not actually exist at a specified location or 2486 if an infant did not actually consume the milk.

2487

RG 1.21, Rev. 3, Page 56 2488 impacted areasThe areas with some reasonable potential for residual radioactivity in excess of natural 2489 background or fallout levels. The NRC discusses impacted areas in 10 CFR 50.2 and 2490 NUREG-1757. For example, impacted areas include locations where radiological leaks or spills 2491 have occurred within the onsite environs (i.e., outside of the facilitys systems, structures, and 2492 components). (See also the definition for significant contamination.)

2493 2494 leachateWater containing contaminants that is percolating downward from a pond or lake into the 2495 subsurface.

2496 2497 less-significant release pointAny location from which radioactive material is released as a liquid or 2498 gaseous effluent contributing less than or equal to 1 percent of the activity discharged from all the 2499 release points for a particular type of effluent considered. RG 1.109 lists the three types of 2500 effluent as (1) liquid effluents, (2) noble gases discharged to the atmosphere in gaseous 2501 radioactive waste, and (3) all other nuclides discharged to the atmosphere in gaseous radioactive 2502 waste.

2503 2504 Example: If 1,000 curies (Ci) of tritium are released in all liquid effluents in a given period of 2505 time (e.g., a typical calendar year or fuel cycle) and 0.01 Ci of tritium is released in steam 2506 generator blowdown, then the steam generator blowdown would be a less-significant release 2507 point. Similarly, for gaseous releases of radionuclides other than noble gases (i.e., iodine, 2508 particulates, and tritium), if the total effluents are 10 Ci (iodine, particulates, and tritium), and the 2509 refueling water storage tank released 0.009 Ci of iodine, particulates, and tritium, then the 2510 refueling water storage tank would be a less-significant release point. In both examples, the 2511 sample frequency can be adjusted to an appropriate frequency for the less-significant release 2512 point. Samples collected from these systems for other programs (e.g., detection of 2513 primary-to-secondary leakage) must still be collected and analyzed at the frequencies specified by 2514 the other programs.

2515 2516 licensed materialSource material, special nuclear material, or byproduct material received, possessed, 2517 used, transferred, or disposed under a general or specific license issued by the Commission.

2518 2519 lower limit of detection (LLD)The a priori smallest concentration of radioactive material in a sample 2520 that will yield a net count, above system background, that will be detected with 95-percent 2521 probability with only a 5-percent probability of falsely concluding that a blank observation 2522 represents a real signal (see NUREG-1301, NUREG-1302, and NUREG/CR-4007, Lower Limit 2523 of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and 2524 Environmental Measurements, issued September 1984 (Ref. 91)).

2525 2526 maximum exposed individualIndividuals characterized as maximum exposed with regard to food 2527 consumption, occupancy, and other usage in the vicinity of the plant site. As such, the maximum 2528 exposed individual represents individuals with habits that are considered to be maximum 2529 reasonable deviations from the average for the population in general. Additionally, in 2530 physiological or metabolic respects, the maximum exposed individual is assumed to have those 2531 characteristics that represent the averages for the corresponding age group in the general 2532 population. (This term typically refers to members of the public.) RG 1.109 contains additional 2533 information.

2534 2535 member of the public (10 CFR Part 20)Any individual except when that individual is receiving an 2536 occupational dose.

2537 2538

RG 1.21, Rev. 3, Page 57 member of the public (40 CFR Part 190)Any individual that can receive a radiation dose in the 2539 general environment, whether the individual may or may not also be exposed to radiation in an 2540 occupation associated with a nuclear fuel cycle. However, an individual is not considered a 2541 member of the public during any period in which the individual is engaged in carrying out any 2542 operation that is part of a nuclear fuel cycle.

2543 2544 minimum detectable concentrationThe smallest activity concentration measurement that is 2545 practically achievable with a given instrument and type of measurement procedure. The 2546 minimum detectable concentration depends on factors involved in the survey measurement 2547 process (surface type, geometry, backscatter, and self-absorption) and is typically calculated 2548 following an actual sample analysis (a posteriori). (See NUREG-1507, Minimum Detectable 2549 Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field 2550 Conditions, issued June 1998 (Ref. 92)).

2551 2552 mixed mode releaseA gaseous effluent release made from a height higher than a ground-level release 2553 but less than an elevated release where, sometimes, because of a lack of plume rise 2554 (e.g., buoyancy, momentum, wind speed), a proper estimate of radionuclide transport and 2555 diffusion requires mathematically splitting the plume into (1) an elevated component and (2) a 2556 ground-level component to properly account for building wake effects, release, or ambient 2557 conditions (or a combination of all three). (RG 1.111 contains further guidance.)

2558 2559 monitoringWith respect to radiation or radiation protection, the measurement of radiation levels, 2560 concentrations, surface area concentrations, or quantities of radioactive material and the use of 2561 results of these measurements to evaluate potential exposures and doses.

2562 2563 nonroutine, planned dischargeAn effluent release from a release point not defined in the ODCM but 2564 that has been planned, monitored, and discharged in accordance with 10 CFR 20.2001 (e.g., the 2565 discharge of water recovered during a spill or leak from a temporary storage tank).

2566 2567 nuclear fuel cycleThe operations defined to be associated with the production of electrical power for 2568 public use by any fuel cycle through the use of nuclear energy (see 40 CFR 190.02).

2569 2570 onsite environsLocation within the site boundary but outside of the systems, structures, or components 2571 described in the final safety analysis report or the ODCM.

2572 2573 operability (operable)The ability of a system, subsystem, train, component, or device to perform its 2574 specified safety function(s) and the ability of all necessary attendant instrumentation, controls, 2575 normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary 2576 equipment (required for the system, subsystem, train, component, or device to perform its 2577 specified safety function(s)) to perform their related support function(s).

2578 2579 principal radionuclideOne of the principal gamma emitters listed in NUREG-1301 and 2580 NUREG-1302, Tables 4.11-1 and 4.11-2, or, from a risk-informed perspective, a radionuclide that 2581 contributes either (1) greater than 1 percent of the 10 CFR Part 50, Appendix I, design objective 2582 dose when all radionuclides in the type of effluent are considered, or (2) greater than 1 percent of 2583 the activity of all nuclides in the type of effluent being considered. RG 1.109 lists the three types 2584 of effluents as (1) liquid effluents, (2) noble gases discharged to the atmosphere, and (3) all other 2585 nuclides discharged to the atmosphere. In this RG, the terms principal radionuclide and 2586 principal nuclide are synonymous since this document is only concerned with measuring, 2587 evaluating, and reporting radioactive materials in effluents.

2588 2589

RG 1.21, Rev. 3, Page 58 radioactive dischargeThe emission of an effluent (i.e., containing plant-related, licensed radioactive 2590 material) into the unrestricted area. (See also the definition for effluent discharge.)

2591 2592 radioactive releaseThe emission of an effluent (i.e., containing plant-related, licensed radioactive 2593 material). (See also the definition for effluent release.)

2594 2595 real exposure pathwayAn exposure pathway in which plant-related radionuclides in the environment 2596 at (or from) a specified location cause exposure to an actual individual. For example, the 2597 grass-cow-milk-infant-thyroid exposure pathway would be considered a real exposure pathway if 2598 the grass, the cow, and the milk actually existed at a specified location and an infant actually 2599 consumed the milk. For purposes of compliance with 10 CFR Part 50, Appendix I, the individual 2600 must be a member of the public.

2601 2602 real individual (10 CFR 72) - Any individual who lives, works, or engages in recreation or other 2603 activities close to the ISFSI/MRS for a significant portion of the year.

2604 2605 release sourceA system, structure, or component (containing radioactive material under the licensees 2606 control) where radioactive materials are contained before release.

2607 2608 release pointA location from which radioactive materials are released from a system, structure, or 2609 component (including evaporative releases and leaching from ponds and lakes in the controlled or 2610 restricted area before release under 10 CFR 20.2001). For release points monitored by plant 2611 process radiation monitoring systems, the release point is associated with the piping immediately 2612 downstream of the radiation monitor. (See also the definition for significant release point.)

2613 Several release sources may contribute to a common release point.

2614 2615 residual radioactivityRadioactivity in structures, materials, soils, groundwater, and other media at a 2616 site resulting from activities under the licensees control. This includes radioactivity from all 2617 licensed and unlicensed sources used by the licensee, but it excludes background radiation. It 2618 also includes radioactive materials remaining at the site as a result of routine or accidental 2619 releases of radioactive material at the site and previous burials at the site, even if those burials 2620 were made in accordance with 10 CFR Part 20.

2621 2622 restricted areaAn area, access to which is limited by the licensee for the purpose of protecting 2623 individuals against undue risks from exposure to radiation and radioactive materials. Restricted 2624 area does not include areas used as residential quarters, but separate rooms in a residential 2625 building may be set apart as a restricted area.

2626 2627 route of exposureA specific path (or delivery mechanism) by which radioactive material can 2628 eventually cause a radiation dose to an individual. The path typically includes a type of 2629 environmental medium (e.g., air, grass, meat, or water) as the starting point and a recipients 2630 organ or body as the end point. For example, the grass-cow-milk-infant-thyroid route of exposure 2631 may contribute to the ingestion exposure pathway. Additionally, several routes of exposure may 2632 contribute to a single exposure pathway.

2633 2634 scaling factorA factor used to estimate the unknown activity of a radionuclide based on its ratio to the 2635 activity of a readily measured radionuclide or other parameter (e.g., carbon-14 scaled to power 2636 generation).

2637 2638 significant contaminationAs used for 10 CFR 50.75(g) recordkeeping, a quantity, concentration, or 2639 both, of residual radioactivity that would require remediation during decommissioning in order to 2640

RG 1.21, Rev. 3, Page 59 terminate the license by meeting the unrestricted use criteria stated in 10 CFR 20.1402 (see 2641 NUREG-1757).

2642 2643 significant release pointAny location from which radioactive material is released that contributes 2644 greater than 1 percent of the activity discharged from all the release points for a particular type of 2645 effluent considered. RG 1.109 lists the three types of effluent as (1) liquid effluents, (2) noble 2646 gases discharged to the atmosphere in gaseous radioactive waste, and (3) all other radionuclides 2647 discharged to the atmosphere in gaseous radioactive waste.

2648 2649 significant residual radioactivitySee the definition for significant contamination.

2650 2651 site boundaryThe line beyond which the licensee owns, leases, or otherwise controls land or property.

2652 2653 site environsLocations outside of the nuclear power plant systems, structures, or components as 2654 described in the final safety analysis report or the ODCM.

2655 2656 solid radioactive waste (solid waste)solid material for which the licensee foresees no further use.

2657 2658 source checkA qualitative assessment of the channel response when the channel sensor is exposed to a 2659 source of increased radioactivity.

2660 2661 standard (instrument or source) (see ANSI N323C-2009 and ANSI N42.22-2006, Traceability of 2662 Radioactive Sources to the National Institute of Standards and Technology (NIST) and 2663 Associated Instrument Quality Control (Ref. 93):

2664 2665 National standarda standard determined by a nationally recognized, competent 2666 authority to serve as the basis for assigning values to other standards of the quantity 2667 concerned. In the United States, this is an instrument, source, or other system or device 2668 maintained and promulgated by the NIST.

2669 2670 Primary standarda standard that is designated or widely acknowledged as having the 2671 highest metrological qualities and whose value is accepted without reference to other 2672 standards of the same quantity.

2673 2674 Secondary standarda standard whose value is assigned by comparison with a primary 2675 standard of the same quantity.

2676 2677 Reference standarda standard, generally having the highest metrological quality 2678 available at a given location or in a given organization, from which measurements made 2679 there are derived.

2680 2681 Transfer standardA standard used as an intermediary to compare standards. (If the 2682 intermediary is not a standard, the term transfer device should be used.)

2683 2684 Working standarda standard that is used routinely to calibrate or check material 2685 measures, measuring instruments, or reference materials. A working standard is usually 2686 traceable to the NIST.

2687 2688 surveyAn evaluation of the radiological conditions and potential hazards incident to the production, 2689 use, transfer, release, disposal, or presence of radioactive material or other sources of radiation.

2690

RG 1.21, Rev. 3, Page 60 When appropriate, such an evaluation includes a physical survey of the location of radioactive 2691 material and measurements or calculations of levels of radiation, or concentrations or quantities 2692 of radioactive material present.

2693 2694 type of effluentA grouping of radioactive releases into one of the three categories listed in 2695 10 CFR Part 50, Appendix I, paragraphs A-C. RG 1.109 classifies the three categories as 2696 (1) liquid effluents, (2) noble gases discharged to the atmosphere in gaseous radioactive waste, 2697 and (3) all other nuclides discharged to the atmosphere in gaseous radioactive waste.

2698 2699 unlicensed materialRadioactive material discharged as licensed material in effluents and background 2700 radioactivity (including global fallout). Licensed radioactive material becomes unlicensed 2701 radioactive material upon discharge in effluents, in accordance with 10 CFR 20.2001.

2702 2703 uncontrolled dischargeAn effluent discharge that does not meet the definition of a controlled 2704 discharge. (See also the definition of controlled discharge.)

2705 2706 uncontrolled releaseAn effluent release that does not meet the definition of a controlled release. (See 2707 also the definition of controlled release).

2708 2709 unplanned dischargeThe unintended or unexpected discharge of liquid or airborne radioactive 2710 material to the unrestricted area. Examples of an unplanned discharge include the following:

2711 2712 the unintentional discharge of a wrong waste gas decay tank (or bulk liquid radioactive 2713 waste tank), or 2714 2715 the failure of a radiation monitor to divert liquid to the radioactive waste system in the 2716 case where radioactivity is present and the automatic alarm/trip function fails to divert 2717 material to liquid radioactive waste and that material (or a portion of that material) 2718 instead discharges to the environment.

2719 2720 unplanned releaseThe unintended or unexpected release of liquid or airborne radioactive material to 2721 the onsite environment. An example of an unplanned release would include a plant occurrence 2722 that results in a leak or spill of radioactive material to onsite areas, requiring a report under 2723 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors, or 2724 10 CFR 50.73, License event report system. (See HPPOS-254, Definition of Unplanned 2725 Release, issued February 1994 (Ref. 94)).

2726 2727 For example, if a licensee has prepared documents describing an intended release (e.g., a 2728 preliminary radioactive waste release permit) in advance of the evolution, and the intended 2729 release occurs as planned, then the release is a planned release. If such documents (e.g., a 2730 preliminary release permit) are not prepared (or considered/evaluated) before the release, it is 2731 potentially an unplanned release (and additional information may be required to determine if it is 2732 an unplanned release).

2733 2734 unrestricted areaAn area for which the licensee neither limits nor controls access.

2735 2736 uranium fuel cycleThe operations of milling of uranium ore, chemical conversion of uranium, isotopic 2737 enrichment of uranium, fabrication of uranium fuel, generation of electricity by a 2738 light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium 2739 fuel, to the extent that these directly support the production of electrical power for public use 2740 using nuclear energy, but excludes mining operations, operations at waste disposal sites, 2741

RG 1.21, Rev. 3, Page 61 transportation of any radioactive material in support of these operations, and the reuse of 2742 recovered nonuranium special nuclear and byproduct materials from the cycle.

2743 2744

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

2749

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

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2938 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.

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2982 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.

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2985 2986

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2991 2992

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

2996 2997

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3000 3001

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3003 3004

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3010 3011

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3013 3014

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3016 3017

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

3019 3020

83.

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3022 3023

84.

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3025 3026

85.

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3028 3029

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3031 3032

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3034 3035

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3038 3039

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3041 3042

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3044 3045

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3048 3049

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3052 3053

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3059

RG-1.21, Rev. 3, Page 69 BIBLIOGRAPHY 3060 3061 U.S. Nuclear Regulatory Commission Documents 3062 3063 NUREG-Series Reports 3064 3065 U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports 3066 for Nuclear Power Plants, NUREG-0800, Section 2.3.5, Long-Term Atmosphere Dispersion Estimates 3067 for Routine Releases, Revision 3, Washington, DC, March 2007.

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3074 3075 Regulatory Guides 3076 3077 U.S. Nuclear Regulatory Commission, Design Guidance for Radioactive Waste Management Systems, 3078 Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, Regulatory 3079 Guide 1.143, Revision 2, November 2001.

3080 3081 U.S. Environmental Protection Agency Documents 3082 3083 U.S. Code of Federal Regulations, National Primary Drinking Water Regulations, Part 141, Chapter 1, 3084 Title 40, Protection of Environment.

3085 3086 National Standards and Industry Reports 3087 3088 ANSI, Performance Criteria for Radiobioassay, ANSI N13.30-1996, New York, NY.

3089 3090 ANSI/ANS, Determining Meteorological Information at Nuclear Facilities, ANSI/ANS 3.11-2005, 3091 New York, NY, January 2005.

3092 3093 ANSI, Calibration and Use of Germanium Spectrometers for the Measurement of Gamma-Ray Emission 3094 Rates of Radionuclides, ANSI N42.14-1999, New York, NY.

3095 3096 ANSI/National Conference of State Legislatures (NCSL), American National Standard for Expressing 3097 UncertaintyU.S. Guide to the Expression of Uncertainty in Measurement, ANSI/NCSL Z540-2-1997 3098 (reapproved 2002), New York, NY.

3099 3100 NIST, Guidelines for Evaluating and Expressing the Uncertainty of NIST Measurement Results, 3101 Technical Note 1297, Gaithersburg, MD, September 1994.

3102 3103 EPRI, Groundwater Monitoring Guidance for Nuclear Power Plants, Report No. 1011730, Palo Alto, 3104 CA, September 2005.

3105 3106 EPRI, Groundwater Protection Guidelines for Nuclear Power Plants, Rev. 1 Report No. 3002000546, 3107 Palo Alto, CA, November 2007.

3108

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

Ci Particulates Ci Tritium Ci Gross Alpha Ci C-14 Ci

RG 1.21, Rev. 3, Appendix A, Page A-2 Table A-1A - Gaseous EffluentsGround-Level ReleaseBatch Mode Fission and Activation Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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/

Halogens UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-1B - Gaseous EffluentsGround-Level ReleaseContinuous Mode Fission and Activation Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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/

Halogens UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-1C - Gaseous EffluentsElevated ReleaseBatch Mode Fission and Activation Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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/

Halogens UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-1D - Gaseous EffluentsElevated ReleaseContinuous Mode Fission and Activation Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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/

Halogens UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-1E - Gaseous EffluentsMixed Mode ReleaseBatch Mode Fission and Activation Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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/

Halogens UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-1F - Gaseous EffluentsMixed Mode ReleaseContinuous Mode Fission and Activation Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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/

Halogens UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-8 Table A Liquid EffluentsSummation of All Releases SUMMATION OF ALL LIQUID RELEASES UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL UNCERTAINTY (%)

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

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

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

(before dilution)

Liters Quarterly Dilution Water Used for Above Liters Average Stream Flow m3/s

RG 1.21, Rev. 3, Appendix A, Page A-9 Table A-2A - Liquid EffluentsBatch Mode Fission and Activation Products UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-10 Table A-2A - Liquid EffluentsBatch Mode (continued)

Dissolved and Entrained Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-11 Table A-2B - Liquid EffluentsContinuous Mode Fission and Activation Products UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-12 Table A-2B - Liquid EffluentsContinuous Mode (continued)

Dissolved and Entrained Gases UNITS QUARTER 1 QUARTER 2 QUARTER 3 QUARTER 4 TOTAL 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-13 Table A Solid Waste and Irradiated Fuel Shipments A. SOLID RADIOACTIVE WASTE SHIPPED FROM THE UNIT (not irradiated fuel)

TYPE OF WASTE NUMBER OF SHIPMENTS VOLUME (m3)

ACTIVITY OF MAJOR NUCLIDES (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-14 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 Dose Limit, Total Body 1.5 mrem 1.5 mrem 1.5 mrem 1.5 mrem 3 mrem Total Body Dose

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

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

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

% of Dose Limit Gaseous Effluent Organ Dose Limit (iodine, tritium, particulates with >8-day half-life) 7.5 mrem 7.5 mrem 7.5 mrem 7.5 mrem 15 mrem 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-15 Table A EPA 40 CFR Part 190 Dose Limits to an Individual in the Unrestricted Area 1

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-16 Table A-6. Supplemental Information 5

6

1.

Abnormal Releases and Abnormal Discharges (e.g., leaks and spills) 7 8

2.

Nonroutine, Planned Discharges (e.g., pumping of leaks and spills for remediation, results of 9

groundwater monitoring to quantify effluent releases to the offsite environment) 10 11

3.

Radioactive Waste Treatment System Changes 12 13

4.

Annual Land Use Census Changes 14 15

5.

Effluent Monitor Instrument Inoperability 16 17

6.

ODCM Changes 18 19

7.

Process Control Program Changes 20 21

8.

Errata/Corrections to Previous ARERRs 22 23

9.

Other (narrative description of other information that is provided to the NRC, such as in the 24 ARERR or ISFSI reports).

25